ML17227A787

Chapter 15 Event Review & Analysis for 25% Steam Generator Tube Plugging

ML17227A787
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Site:	Saint Lucie
Issue date:	02/28/1993
From:	Duggan K SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
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ML17227A785	List: L-93-035, Application for Amend to License DPR-67,revising TS Figure 2.1-1,Tables 2.2-1 & 3.2-1 & Bases Pages B2-1 & B2-2 for Reduction of RCS Design Flow to Accomodate Future SG Tube Plugging Up to 25% (Average) ML17227A786 ML17227A787 ML17227A788 ML17227A789
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EMF-92-165, NUDOCS 9303310058
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Siemens Power Corporation - Nuclear Division EMF-92-1 65 Issue Date:

p g g3 St. Wcle Unit 1 Chapter 15 Event Review and Analysis for 25%%d Steam Generator Tube Plugging Prepared by:

M. Ou, Engineer PWR Reload Analysis PWR Nuclear Engineering February 1993 sp 93033i0058 '9303i9 PDR

.,ADOCK 05000335 P

'PDR

A DUS 0

I Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the Agreement between Siemens Power Corporatke and the Customer pursuant to whkh this document Is issued.

Accordingly, except as otherwise expressly provided in such Agreement, neither Siemens Power Corporation nor any person acting on its behalf makes any warranty or representation, expressed or Implied, with respect to the accuracy, completeness, or usefulness of the informalion contained in this document, or that the use of any lnformathn, apparatus, method or process disclosed In this document wINnot Infringe privately owned rights; or assumes any liabilitieswith respect to the use of any information, apparatus, method or process disclosed In this document.

~ Informathn contained herein is for the sole use of the Customer.

In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions whkh may be included in the information contained In this document, the recipient, by its acceptance ofthis document, agrees not to publish or make public use (in the patent use of the term) of such information until so authortzed In wrNng by SIemens Power Corporation or until after six (8) months followingtermination or expiration of the aforesakf Agreement and any extension thereof, unless expressly provided in the Agreement.

No rights or licenses in or to any patents are implied by the furnishing of this document.

Table of Contents EMF-92-165 Page i

Section Pacae

1.0 INTRODUCTION

2.0

SUMMARY

2-1 3.0 CHAPTER 15 EVENT REVIEW....,.....

3-1 3.1 Summary 3.2 Review of Events

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3-2 4.0

SUMMARY

OF OPERATING LIMITS 4-1 4.1 Reactor Protection System 4.2 Specified Acceptable Fuel Design Limits 4.3 Setpoint Analysis.. ~........

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4-1 5.0 LOSS OF EXTERNALLOAD

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5 1 5.1 Event Description 5.2 Definition of Events Analyzed 5.3 Analysis Results

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5-1 5-2 6.0 REACTOR COOLANT PUMP ROTOR SEIZURE

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6-1 6,1 Event Description..

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6.2 Definition of Events Analyzed 6.3 Analysis Results

7.0 REFERENCES

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List of Tables EMF-92-165 Page ii Table 3.1 4.1 4.2 4.3 4,4 4.5 5.1 5,2 Summary of St. Lucie Unit 1 Chapter 15 Event Review For This Analysis Uncertainties Applied in LSSS Calculations..

Uncertainties Applied in the TM/LP LSSS Calculation Additional LSSS Trip Functions...

Uncertainties Applied in the LCO Calculations Uncertainties Applied in DNB LCO Calculations Summary of Events for the Loss of External Load Nomenclature Used in Plotting Results

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Pacae 3-18 4-5 4-6 4-7 4-8

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List of Fi ures EMF-92-1 65 Page iii Ficiure Pacae 4.1 4.2 4.3 5.1 5.2 5.3 5.4 5.5 5.6 St. Lucie Unit 1, Thermal Margin/low Pressure Setpoint, Part 1

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St. Lucie Unit 1, Thermal Margin/low Pressure Setpoint, Part 2 (QR1)

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Verification of Cycle 11 DNB Limiting Condition of Operation (with a Technical Specification Flow of 355,000 GPM).......

Reactor Power Level for Loss of External Load..

Core Average Heat Flux for Loss of External Load Reactor Coolant System Temperatures for Loss of External Load 0

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Pressurizer Pressure For Loss of External Load Reactivities for Loss of External Load Secondary Pressure for Loss of External Load.........

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4-1 0 4-11 4-12 5-5 5-6 5-7 5-8 5-9 5-10

Ust of Abbreviations EMF-92-165 Page iv Abbreviation Definition AOO ADV ASGPTF ASI CE CEA CEDM Anticipated Operational Occurrence Atmospheric Dump Valve Asymmetric Steam Generator Pressure Trip Function Axial Shape Index Combustion Engineering Control Element Assembly Control Element Drive Mechanism DNB DNBR ECCS EOC FSAR HP LCO LOCA LOFA LOEL LPD LSSS MDNBR MSIV MTC PCS PCT PORV PZR RPS RTD Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Emergency Core Cooling System End of Cycle Final Safety Analysis Report High Pressure Limiting Condition of Operation Loss of Coolant Accident Loss of Feedwater Accident Loss of External Load Local Power Density Limiting Safety System Setting Minimum Departure from Nucleate Boiling Ratio Main Steam Isolation Valve Moderator Temperature Coefficient Primary Coolant System Peak Clad Temperature or Percent Power Operated Relief Valve Pressurizer Reactor Protection System Resistance Temperature Device

List of Abbreviations (Continued)

EMF-92-165 Page v Abbreviation SAFDL SG SGSRV SGTP SPC SRP TM/LP Tr UFSAR VHPT Definition Specified Acceptable Fuel Design Limit Steam Generator Steam Generator Safety Relief Valve Steam Generator Tube Plugging Siemens Power Corporation - Nuclear Division Standard Review Plan Thermal Margin/Low Pressure Turbine Trip Updated Final Safety Analysis Report Variable High Power Trip

EMF-92-165 Page 1-1 St. Lucle Unit 1 Chapter 15 Event Review and Analysis for 25% Steam Generator Tube Plugging 1.0 INTROOUCTlON This analysis supports a Steam Generator Tube Plugging (SGTP) level of 25% with an asymmetry of +/- 7% and a minimum Technical Specmcation limiton Primary Coolant System (PCS) flowof 355,000 gpm.<'>

The remainder of the conditions used for this analysis are based on those reported for St. Lucie Unit 1 Cycle 11, reported in Reference 2.

The topics addressed in this report include a review of the SRP Chapter 15 Events, Summary of Operating Limits (Setpoints),

Loss of External Load analysis (PCS pressurization event) (15,2.1), and Reactor Coolant Pump Rotor Seizure analysis (15.3.3). The, Large Break Loss of Coolant Accident (LOCA) (15.6.5) and the Small Break LOCA (15,6.5) analyses are reported in references 3 and 4, respectively.

EMF-92-165 Page 2-1 2.0

SUMMARY

A review of SRP Chapter 15 events was performed to assess the impact of an increase in SGTP of 25%, with an asymmetry of +/- 7% and a reduced minimum Technical Specification PCS flow of 355,000 gpm. The events identmed that require reanalysis are listed below. This review uses Cycle 11 as a basis.(

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EVENT REQUIRING ANALYSIS EVENT DESCRIPTION 15.2.1 Loss of External Load SECTION OF REPORT 5.0 SECTION DESCRIPTION Loss of External Load 15.3.1 Loss of Forced Reactor Coolant Flow 4.3 Setpoint Analysis 15.3.3 Reactor Coolant Pump Rotor Seizure 15.4.3 CEA Misoperation (Dropped CEA Only) 15.6.5 Large Break LOCA 15.6.5 Small Break LOCA 6.0 4.3 NOTE 1 NOTE 2 Reactor Coolant Pump Rotor Seizure Setpoint Analysis NOTE 1:

NOTE 2:

The large break LOCA analysis is reported in reference 3.

The small break LOCA analysis is reported in reference 4.

To insure DNB criteria is still met with the increased SGTP and the decreased Technical Specmcation PCS flow, the TM/LPand DNB LCO setpoint analyses were reanalyzed in this report (Section 4.3). The TM/LPtrip setpoint analysis shows an excess margin of protection is provided by the existing trip equation. The validity of the existing DNB LCO barn for allowable core power as a function of ASI was vermed to ensure adherence to the SAFDL on DNB during CEA drop and Loss'-Flow Anticipated Operational Occurrences (AOOs).

The Loss of External Load (LOEL) transient reanalysis demonstrates that both the primary and secondary system pressure relief capacities are sufficient to limitthe pressures of both systems to less than 110% of their design limits (Section 5.0).

EMF-92-165 Page 2-2 The Reactor Pump Rotor Seizure event was reanalyzed to assess the event's challenge to the DNB limit (Section 6.0). The analysis shows that 1% of the rods experience DNB for this event.

The reference radiological consequences analysis<'

assumed 2.5% of the rods experienced DNB.

Therefore, the reference analysis results bound operation with increased SGTP and reduced PCS flow rate,

EMF-92-165 Page 3-1 3.0 CHAPTER 15 EVENT REVIEW I

A review of the Chapter 15 events for St. Lucie Unit 1 was performed to support an increase in SGTP to 25% with an asymmetry of +/- 7% and a Technical Specification PCS flow of 355,000 gpm.('> The remainder of the conditions assumed in this analysis are the same as those used in Cycle 11.< )

There are no changes in the fuel design, core physics parameters, or plant operating procedures assumed for this analysis, The changes in Technical Specifications and plant hardware considered in the event review are described in Section 3.1.

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~Summa Following is a list of the changes considered in this event review:

Original This Parameter 'cte t t Reviewt i Steam Generator Tube Plugging (SGTP)

Technical Specwcatlon PCS Flow Rate Limit Reactor Coolant Flow Rate Uncertainty SG secondary pressure 15%

370,000 gpm 11,100 gpm 807 psia 25%

355,000 gpm 14,500 gpm 790 psia Based on this review of the SRP Chapter 15 events, the following events were determined to require reanalysis.

All other events are either bounded by another event, bounded by existing analyses of record, or are not in the licensing basis.

1.

15,2.1 Loss of External Load PCS pressurization event.

2.

15.3.1 Loss of Forced Reactor Coolant Flow (4 Pump Coastdown)

Analyzed for DNBR in the DNB-LCO verwcation.

3.

15.3.3 Reactor Coolant Pump Rotor Seizure.

EMF-92-165 Page 3-2 4.

15,4.3 CEA Misoperation (Dropped CEA Only)

Analyzed for DNBR in the DNB-LCO verification.

5.

15.6.5 Large Break LOCA 6.

15.6.5 Small Break LOCA 3.2 Review of Events The events are presented in the same order as that used in the SRP.

A cross reference of SRP and FSAR event enumeration is presented in Table 3.1.

Chapter 15 of the UFSAR and the analyses performed for the Cycle 6 plant transient licensing submittal<

> are taken as the licensing basis forthe review. Event acceptance and single failure criteria are those presented in the FSAR for each event.

No change in event initiators were considered in the review process.

The methodology of the review is as follows:

1)

Assessed the relative margin of similar events to event acceptance criteria by examining the controlling parameters or event initiators.

2)

Determined the bounding event of similar events based on these margins.

3)

Examined the effect on the limiting event due to the increased SGTP and the Decreased Technical Specwcatlon PCS flow.

4)

Determined whether the event was bounded by an accepted event analysis.

5)

Reanalyzed the limwng event when it was not bounded by an accepted event analysis.

This section of the report presents a description of the review and disposition of each Chapter 15 event.

Each section includes a brief discussion of the event and the conclusion of the review. The parenthetical numbers in the section heading denote the SRP event number.

summary of the results of the event review is presented in Table 3.1,

EMF-92-165 Page 3-3 3.2.1 Increase ln the Heat Removal b the Seconda S stem (15.1)

The events in this category of events were evaluated by calculating the increase in system cooling due to the event initiator. Reference 2 dispositioned the AOOs in this category to be bounded from the standpoint of DNBR by the Increase in Steam Flow event (15.1.3).

The increase in SGTP and decrease in Technical Specification reactor coolant flow will reduce the transient heat transfer rate from the primary to the secondary systems.

The changes in SGTP and flow will affect all of the events in this category in a very similar manner.

Therefore, the increase In steam flow (15.1.3) remains the bounding event.

Only Event 15.1.3 is discussed relative to DNBR.

3.2.1.1 Increase in Steam Flow (15.1.3) e Three event Initiators are postulated:

1)

Malfunction of the generator load limiter, resulting in about 10% increase in load due to opening of the turbine admission valves.

2)

Opening of the steam dump and bypass valves at power due to turbine trip permissive failure, resulting in an increase to 143.4% of rated load.

3)

Opening of the steam dump, and bypass valves at hot standby due to controller malfunction.

The primary coolant system response forthis event is similar to the feedwater malfunction events.

The steam demand increase results in depressurizatlon of the steam generators and the consequent cooldown of the primary coolant system.

The subevents are addressed individually below.

Subevent 1:

The magnitude of the primary coolant cooldown for this event is limited by the capacities of the admission valves and the high pressure turbine admission nozzle.

The steam capacity is not more than 110% of rated, so the cooling load increase for this event was conservatively taken to be 10% of full load.

This Is much less than the steam flow increase in Subevent 2, and is therefore bounded by Subevent 2.

EMF-92-165 Page 3-4 Subevent 2:

This event was analyzed for Cycle 6 for the first reload of SPC fuel.< ) The analysis showed that from the standpoint of DNB, this event is bounded by the loss of "oolant flow event.

The increase in SGTP and decrease in Technical Specification PCS flowwill:end to reduce the calculated MDNBRs forall the DNB events in a similar manner. Therefore, this event willcontinue to be bounded by the loss of coolant flow event.

Reanalysis of the event is not required.

T Subevent 3:

The UFSAR (Table 15.2.11-4) shows that the reactor trips on the VHPT at 44.6 seconds and 40% power.

Due to the low power level, the MDNBR for this subevent is much greater than for Subevent 2. Event response is therefore bounded by that for Subevent 2.

3.2.1.2 Inadvertent 0 enln of a Steam Generator Relief or Safe Valve (15.1.4),

The event is evaluated in the FSAR to assess radiological consequences.

Radiological effects for this event are based on releases due to opening a power operated atmospheric dump valve.

The radiological consequences are based on primary and secondary coolant activity and primary to secondary leak rate Technical Specification limits which remain unchanged from those used in the reference radiological analysis.<'

Therefore, the reference analysis remains bounding for Cycle 12. The radiological consequences are independent of burnup or neutronic desic.'.

3.2.1.3 Steam S stem Pi ln Failures Inside and Outside of Containment (15.1.5)

This event is initiated by complete severance of a main steam pipe.

Automatic closure of the MSIVs stops blowdown of the intact steam generator a few seconds after event initiation.

Secondary blowdown and depressurizatlon cause a cooldown of the moderator, resulting in an erosion of shutdown margin and possible return to power at EOC conditions.

The event was analyzed in Reference 8 for Cycle 6 and on a confirmatory basis using advanced analytical methods in Reference 10, The referenced analysis<

> conservatively assumes no SGTP.

Increased tube plugging would reduce the cooldown rate of the PCS and subsequer 'eturn to power. A decrease in Technical Specification PCS flow rate does not affect this event because the limiting DNB case occurs with a loss of offsite power and PCS flows are governed by natural circulation. Consequently, the analysis of record is bounding.

Reanalysis is not required.

EMF-92-1 65 Page 3-5 Radiological consequences of this event are bounded by that of the LOCA (15.6.5).

3.2.2 Decrease In Heat Removal B The Seconda S stem The events in this category result in a rapid pressurization of the steam generator.

The pressurization results in a temperature rise in the steam generator secondary side and in the primary coolant system.

At the beginning of cycle when the MTC is conservatively assumed to be positive, reactor power increases.

The transient is usually terminated by either the high pressure or TM/LP trip. However, the primary temperature and power rise can potentially result in violating either the 110% over-pressurization limit or the SAFDL on DNB.

Several assumptions are made for the analysis of the events in this category.

The assumptions are:

The LOEL event is Initiated with turbine stop valve closure, a fast acting (0.1 to 0.3 sec.

closure) valve.

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The steam bypass and dump systems are assumed inoperative.

Of the potential initiators, the assumption of the closure of the turbine stop valve is most limiting since it results in the most rapid decrease in steam flow.

Other events in the 15.2 category of events are Turbine Trip (T7) (15.2.2), Loss of Condenser Vacuum (1 5.2.3), and Closure of Main Steam Isolation Valve (MSIV) (1 5.2.4), The use of the stop valve without concurrent reactor trip in the LOEL event bounds the TT event since reactor trip on turbine trip would be expected.

An earlier reactor trip results in more margin to the acceptance criteria.

The loss of condenser vacuum both disables steam bypass and results in a gradual decrease in steam flow compared to that resulting from the stop valve closure.

Since both effects are assumed for the LOEL event, the loss of condenser vacuum event is bounded by the LOEL event.

The MSIVclosure time is greater than turbine stop valve closure time, The more

EMF-92-165 Page 3<

rapid closure of the stop valves produces a more severe system transient than does the MSIV closure.

Therefore, the MSIV closure event is bounded by the LOEL event.

Since the LOEL is the bounding event, it is the only event requiring a review.

3,2.2.1 oss of External Load (15.2.1)

There are two suMvents in the LOEL event:

PCS pressurization and DNB. This event was analyzed for Cycle 6 for the first reload of SPC fuel.<

The analysis showed that from the standpoint of DNB, this event is bounded by the loss of coolant flow event.

The increase in SGTP and decrease in Technical Specification PCS flow will tend to reduce the calculated MDNBRs for all of the DNB events in a similar manner.

Therefore, this event will continue to be bounded by the toss of coolant flow event.

Reanalysis of the event is not required.

Secondary system pressures increase rapidly after the event initiation due to the elimination of steam flow to the turbine. The pressure reaches the setpoint of the secondary safety valves at about 5 seconds and the safety valves open.

Peak secondary pressure is reached at about 7

seconds, and then decreases slowly as the load on the steam generators decreases with decreasing core power. The increased SGTP causes initial SG pressures to be lower, which allows a larger change in pressure during the transient to reach the safety valve setpoints.

Increased SGTP will increase the PCS coolant insurge into the pressurizer because the larger change in secondary pressure and temperature causes a larger increase in the temperature and volume of the PCS liquid. In addition, increased SGTP will reduce the rate of heat transfer from the PCS to the secondary, once the secondary safety valves lift, which may cause more of an overshoot in primary pressure after the opening of the PORV. Both of these effects willtend to produce a higher peak PCS pressure.

Therefore, this event is reanalyzed for increased SGTP for the peak pressure event ln section 5.0 of this report.

Radiological effects for this event are based on primary and secondary coolant activity and primary to secondary leak rate Technical Specification limitswhich remain unchanged from those used in the UFSAR analysis.

Therefore, the UFSAR analysis()

remains bounding.

EMF-92-165 Page 3-7 3.2.2.2 Loss of Nonwmer en AC Power to the Station Auxiliaries (15.2.6)

The toss of AC power to the station auxiliaries event results in:

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Loss of Primary Coolant Pumps

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Loss of Normal Feedwater Auto Diesel Generator Start Release of Control Element Drive Mechanism (CEDM) Holding Coils Turbine Trip - Load Loss The release of the CEDM holding coils shuts the reactor down by inserting the CEAs.

The UFSAR does not take credit for this scram and relies on the low flow trip signal occurring 1

second into the event, The CEAs are released after a delay of 1.15 seconds after the scram signal.

The early part of the event (0-10 seconds) would be the same as the four-pump coastdown event (15.3.1) because the SG inventory would not have been reduced sufficiently to affect heat removal. Therefore, the DNB SAFDLand over-pressurization limitare bounded by the four-pump coastdown event.

In the longer term, the results are bounded by the loss of normal feedwater flow event (15.2.7) since the reactor trip would be delayed further In that event, occurring on the low SG liquid level trip or the high primary system pressure (HP) trip. Therefore, this event is principally analyzed to evaluate the radiological consequences, The decrease in Technical Specification's PCS flow rate does not affect the radiological consequences of this event.

Radiological releases are slightly less severe with increased SGTP for this event.

The initial secondary pressure will be lower such that it takes longer to reach the atmospheric dump valve (ADV) and safety valve setpoints.

So, the total radiological release is less.

Also, the Technical Specification limits on primary and secondary coolant activity and primary to secondary leak rate are unchanged from the reference analysis.<

) Therefore, the UFSAR analysis remains bounding for increased SGTP.

EMF-92-165 Page 3-8 3.2.2.3 Loss of Normal Feedwater Flow (15.2.7)

The event is initiated by a malfunction of the feedwater system, resulting in the loss of normal feedwater flow. This results in a gradual reduction of SG inventory.

The inventory reduction would eventually cause a decrease in heat removal rate from the PCS because of the drop in secondary liquid level below the SG tube bends.

This effectively decreases the heat transport area since the film heat transfer coefficient on the secondary side would be severely reduced in a steam atmosphere.

The reduction in heat removal rate would result in a heatup of the PCS.

Several reactor trips are available to mitigate the consequences of this event.

The event is evaluated for challenge to the primary vessel pressure limit and to assure the adequacy of steam generator inventory. The event is not limiting with respect to DNBR. From the standpoint of primary system pressurization, the LOEL event causes a much more rapid increase in primary system temperatures and pressures because the decrease in heat transfer is more rapid, The loss of normal feedwater flow event is therefore bounded by the LOEL event.

The Loss'-Feedwater event was reanalyzed relative to steam generator inventory in Reference 11.

That analysis conservatively assumed no SGTP, which maximizes the heat transfer rate between the primary and the secondary.

No. SGTP also results in the highest initial secondary

pressure, which maximizes the release of mass from the secondary as a function of time.

Therefore, increased SGTP is bounded by the analysis of record.

A decrease in the Technical Speciflcation PCS flow rate will have negligible effect on the minimum SG Inventory event. Areduced flowrate would result in a slightly lower convective heat transfer coefflcient on the primary side of the SG tubes, which would tend to reduce the severity of the event.

However, this resistance to heat transfer is very small compared to the resistance across the tubes. The differenc in heat transfer rates between the primary and Secondary would be negligibly small. Therefore, the analysis of record remains bounding for the increased SGTP level and the decreased Technical Specification PCS flow rate.

EMF-92-165 Page 3-9 3.2.2.4 Feedwater S stem Pi e Breaks inside and Outside Containment (15.2.8)

This event is a cooidown event In the licensing basis for the plant. As such, the feedwater pipe break event is bounded by the steamline break event since the area for flow in a broken feedwater pipe is less than that of a severed steamline.

The smaller area for flow results in a lower steam relief rate which produces a more benign event.

3,2.3 Decrease in Reactor Coolant S stem Flow Rate The events in this category are initiated by a malfunction of the PCS coolant circulation pumps which reduces reactor coolant flow. The decrease in coolant flow rate also causes an increase in core average temperature.

The decrease in PCS coolant flow and the increase in average core temperature results in an erosion of DNB margin.

3.2.3.1 Loss of Forced Reactor Coolant Flow Four-Pum Coastdown (15.3.1)

The original flow coastdown curve was calculated with a Technical Specification PCS flow rate of 370,000 gpm. The reduction in Technical Specification PCS flow rate to 355,000 gpm, a 4.1%

decrease, would not change the flow coastdown rate significantly, The coastdown algorithm currently used in the setpoint analysis is satisfactory for increased SGTP.

In addition, an uncertainty in the coastdown rate is included in the statistical setpoint analysis. Thus, no system calculations need be performed.

The event was analyzed for the increased SGTP and the reduced Technical Specificatio PCS flowby verifying the DNB LCO barn (Section 4.3.3.1). Axial peaking and the scram reactivity insertion curves for Cycle 11 were used in this analysis.

3.2.3.2 Reactor Coolant Pum Rotor Seizure (15.3.3)

J The event is assumed'to be Initiated by the instantaneous seizure of one of the primary coolant pump shafts.

The resulting heatup transient is treated as either a DNB subwvent or a pressure subwvent.

EMF-92-1 65 Page 3-10 Pressure Subwvent The event was evaluated in Reference 8 and shown to be less limiting with respect to system pressurization than Event 15.2.1, Loss of External Load.

Neither the increased SGTP nor the corresponding Technical Specwcatlon PCS flow rate willsignificantly affect the outcome of this event.

DNB Subwvent The same arguments regarding the rate offlowcoastdown with increased SGTP and a decrease in the Technical Specwcation PCS flowrate for event 15.3.1 (Loss of Reactor Coolant Flow) apply for this event.

However, a decrease in the Technical Specification PCS flow rate will result in lower ONBRs and a higher percentage of fuel rod failures for this event. Therefore, this event will require reanalysis.

The reanalysis of this event is given in section 6.0.

Radiological consequences for high burnup fuel for this event are addressed in Reference 12 3.2.3.3 Reactor Coolant Pum Shaft Break The event is not in the UFSAR, and therefore, not part of the licensing basis.

3.2.4 Reactlv'nd Power Distribution Anomalies 3,2,4.1 ntrolled CEA Withdrawal from a Subcritical or Low Power Startu Condition (15.4.1)

The event is analyzed ln Section 15.2.2 of the St. Lucie Unit 1 UFSAR. The event was initiated from hot zero power condition (532'F, shutdown rods out, 2225 psia) by rod withdrawal from a very low initial power level. Slgnwcant power did not result until neutron multiplication reached a level corresponding to approximately 1% power with the rapid reactivity insertion characteristic of this event.

A rapid power increase occurred, first limited by negative Doppler feedback, then

EMF-92-1 65 Page 3-11 terminated by the VHPT. The peak neutron power was high (150% rated), but the time from trip to MDNBRwas short (2 seconds), resulting in relatively low surface heat flux and consequently, a relatively benign MDNBR. The event is bounded with respect to MDNBR by event 15.4.2, Uncontrolled CEA Withdrawal at Power.

3.2.4.2 Uncontrolled CEA Withdrawal at Power (15.4.2)

The Fast CEA Withdrawal event was analyzed for Cycle 6 in Reference 8. This event is bounded with respect to MDNBR by the Losswf-Coolant Flow event.

The increased SGTP and the decreased Technical Specification PCS flow rate changes do not affect the relative MDNBRs between events, so only the limiting event with respect to DNB needs to be analyzed.

The Loss-of-Coolant Flow event was reanalyzed with the decreased Technical Specification PCS flow, and the increased SGTP in the DNB LCO analysis, Section 4.3.3.1.

Th'erefore, no reanalysis of the Fast CEA WIthdrawal event is required.

Slower CEA withdrawals may terminate on the TM/LPtrip. The TM/LPsetpoint provides designed protection of DNBR limits.

The adequacy of the current TM/LP trip to protect the fuel is addressed in the TM/LPverification analyses for both the decreased Technical Specification PCS flow and the increased SGTP (Section 4.3.1.2). The Slow CEA Withdrawal event was reanalyzed for Cycle 11 to evaluate the effect of increasing the RTD response time from 8 sec to 16 sec.

Neither the increase in SGTP nor the decrease in Technical Specification PCS flowrate adversely affect the outcome of this analysis.

Therefore, reanalysis is not required.

3.2.4.3 CEA Miso eration (15.4.3)

The SRP defInes four subwvents:

1)

Static Misalignment of CEAs 2)

Single CEA Withdrawal

EMF-92-165 Page 3-12 3)

Dropped CEA Bank P

4)

Dropped CEA The first three sub-events are not reported in the UFSAR or Reference 8, and therefore, are not part of the plant licensing basis.

The dropped CEA event produces changes in radial power distribution which are accounted for in the analysis by a radial peaking augmentation factor. The augmentation factor calculated for Cycle 11 is bounded by the value used in the Cycle 6 analysis.<

> The reactor returns to the initial power at EOC by cooling down, since St, Lucie Unit 1 has no automatic rod withdrawal logic. The average coolant temperature which produces a return to 100% power is detei'mined by the moderator and Doppler feedback.

The PCS pressure is determined by the average coolant temperature, Because bounding neutron kinetics parameters are unchanged for Cycle 11 from those employed in the Cycle 6 reference analysis, plant response would not change from that reported for Cycle 6. This event is reanalyzed in the vermcation of the DNB LCO setpoint (Section 4,3.3.1). The increase in SGTP and corresponding decrease in Technical Specification PCS flow are included in this reanalysis.

3.2.4.4 Startu of an inactive Loo (1 5.4.4)

Part loop operation is not permitted by the plant Technical Specification.

Therefore, analysis of this event Is not necessary.

3.2.4.5 Malfunctions that Result in a Decrease ln the Boron Concentration in the The event can occur during all modes of operation.

The modes considered in the UFSAR and Cycle 6@ analyses for this event are:

~

Power (Mode 1)

~

Startup (Mode 2) 0

EMF-92-165 Page 3-13

~

Hot Standby (Mode 3)

~

Hot Shutdown (Mode 4)

~

Cold Shutdown (Mode 5)

~

Refueling (Mode 6)

The SRP acceptance criteria require protecting the fuel SAFDLs for the at power sub-event.

The FSAR presents a verbal analysis which concludes that the TM/LP trip, LPD LSSS and VHPT protect the reactor against this event.

Further, because of the available alarms and indications, there is ample time and information available to allow the operator to take corrective action.

Protracted erroneous dilution is improbable.

This disposition was retained for Cycle 11 and is also retained forthe increase SGTP level and corresponding decrease in Technical Specification PCS flow rate. Also, the acceptance criteria was determined to be satisfied for Modes 2 through 6 when assuming a depiction of B-10 isotope concentration at a rate of 0.1% per effective full power month.

The acceptance criterion for Modes 2 through 6 is that the time to critically allows the operator to terminate the event. The time to criticalitywas reanalyzed for Modes 2 through 6 for Cycle 11 and also examined for this analysis.

Since Mode 6 only considers the mass inventory in the reactor vessel, the increase in SGTP does not affect Mode 6, The affect of the increased SGTP for Modes 2 through 5 was not enough to erode the existing margin. Therefore, the calculated times to lose the required shutdown margin for Modes 2 through 6 were all greater than acceptance criteria.

3.2.4.6 Inadvertent Loadin and 0 eratlon of a Fuel Assembl In An lm ro er Location (1 6.4.7)

The UFSAR presents two suMvents:

~

Misloaded Fuel Pellets or Fuel Rod(s)

~

Erroneous Placement or Orientation of Fuel Assemblies

EMF-92-165 Page 3-14 The discussion presented in the UFSAR is applicable to operation with the increased SGTP and the decreased Technical Specification PCS flow.

3.2,4.7 S ectrum of CEA E ection Accidents (15.4.8)

A control rod ejection accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a control element assembly (CEA) and drive shaft.

The consequence ofthis mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

The rod ejection accident was evaluated with the procedures'developed in the Generic Rod Ejection Analysis(

) for Cycle 11 in Reference 2. Increased SGTP has no significant affect on this event. Prediction of fuel failure is based on fuel centerline melt criteria (or deposited energy),

not on DNB criteria. Therefore,a reduction in Technical Specification PCS flowwillalso not affect this event.

No reanalysis is necessary.

3.2.5 Increase i

Reactor Coolant Invento 3.2.5.1 Inadvertent 0 eration of ECCS (1 5.5.1)

The event is not in the UFSAR and therefore not part of the licensing basis for the plant.

3.2.5.2 CVCS Malfunction that increases Reactor Coolant Invento (15.5.2)

The event is not in the UFSAR and therefore not part of the licensing basis for the plant.

EMF-92-1 65 Page 3-15 3.2.6 Decrease in Reactor Coolant Invento 3.2.6.1 Inadvertent 0 enin of a Pressurizer Pressure Relief Valve (15.6.1)

The event was analyzed for Cycle 6 in Reference 8 as a PCS depressurization event.

The event is bounded with respect to DNBR by the Losswf-Coolant flow event (Section 3.2.3.1). The effect of the increased SGTP and corresponding decrease in Technical Specification PCS flow is considered for that event in the setpoint analyses (Section 4.3).

This event is also used to determine a bias term for the TM/LP trip. The changes being addressed in this disposition do not affect the bias term previously determined for the event.

3,2.6.2 Radiolo ical Conse uences of the failure of Small Lines Car in Prima Coolant Outside Containment (1 5.6.4)

The event is not in the UFSAR and therefore not part of the licensing basis of the plant.

3.2.6.3 Radiolo ical Conse uences of Steam Generator Tube failure (15.6.3)

According to the UFSAR, primary to secondary break flow is choked throughout most of this transient. There are only 400 seconds (from about 600 to 1000 seconds) of the transient where the flow is unchoked; The flow through the broken SG tube is choked before reactor scram.

Therefore, the reduction in secondary pressure due to increased SGTP during this time doesn' affect the amount of primary to secondary leakage.

After reactor scram, the peak pressure reached by the primary is dictated by the opening of the steam dump valves and bypass valves.

Once these valves are opened, the SG pressure response is controlled by the mass and energy balance through the break. Therefore, the overall response of the transient is unaffected by the increased SGTP and the event is bounded by the analysis of record, Reference 12 addresses the radiological consequences for increased burnup fuel.

The decrease in Technical Specification PCS flow does not affect this event.

No reanalysis is required.

EMF-92-165 Page 3-16 3.2.6.4 Loss of Coolant Accident LOCA (15.6.5)

Lar e Break Both the increase in SGTP and the reduction in the Technical Specicatlon PCS flow rate dictate

'his event be reanalyzed.

Increased SGTP affects steam binding in the reflood period of the transient.

The reanalysis of this event is report in Referenc Radiological consequences for increased burnup fuel for this event are provided in Referer

! 2.

Small Break Increased SGTP causes a reduction in the initial PCS coolant inventory, which could cause a

deeper core uncovery and higher PCT.

Also, the timing of the two~hase flow at the break, depressurization rate, and loop seal clearing could be affected by increased SGTP. The analysis for this event ls reported in Reference 4.

A decrease in the Technical Specicatlon PCS flow rate will not affect the SBLOCA event.

The PCS pumps coast down early in the transient such that the initial loop flow rate does not affect the heatup portion of the event.

3,2.7 Radioactive Releases from a S stem or Com onent None of the events described in Section 15.7 of the SRP are affected by an increase in SGTP level or the corresponding decrease in Technical Specification PCS flow. None of these events occur as a 'direct consequence of operation of the reactor.

The conclusions presented in Reference 12 for extended bumup:levels are bounding for an increase in SGTP level and decrease in Technical Specication PCS flow.

EMF.-92-165 Page 3-17 3.2.8 As mmetric E ents The UFSAR includes four events which are initiated by malfunctions that affect only one SG. The St, Lucie Unit 1 plant has an ASGPTF which is set at 135 psi, Thus, if the pressure difference between two SGs exceeds 135 psi, a reactor trip signal occurs.

The UFSAR analysis shows that this trip completely mitigates the consequences of such malfunctions.

The reason for this conclusion is that a trip signal is generated long before any core inlet asymmetries can develop. The decrease in Technical Specification PCS flow does not affect these analyses since the ASGPTF trips the reactor before asymmetries can develop. Thus the symmetric events remain bounding.

This disposNon assumes an increased asymmetric steam generator tube plugging of 32% in one SG and 18% In the other SG. The power split would be about 54% power in the 18% plugged steam generator and 46% power in the 32% plugged steam generator.

Since the SGs are manifolded together, the difference in pressures in the steam generators due to differences in pressure drops from the steam domes to the manifold would be on the order of a few psi. The small difference in pressures between the two steam generators relative to the 135 psid required for a trip would not negate conclusions stated in the UFSAR regarding Asymmetric Events.

Table 3.1 Summary of St. Lucle Unit 1 Chapter 15 Event Review for this Analysis Cross-Reference Table SRP Event/FSAR Event SRP Event ¹ 15.1.1 15.1.2 15.1.3 15.1.4 UFSAR Event ¹ 15.2.10 15.2.10 15.2.11 15.2.11 Event Name Decrease in Feedwater Temperature Increase in Feedwater Flow Increase ln Steam Flow

~ois osition Bounded by 15.1.3 Bounded by 15.1.3 Bounded by Cycle 6 (Plant Response)

Inadvertent.Opening of a Steam Genera-Bounded by 15.1.3, tor Relief or Safety Valve Radiological Consequences bounded by UFSAR 15.1.5 1 5.4.6 Steam System Piping Failures Inside and Outside of Containment Bounded by Cycle 6 Analysis and Reference 10 15.2.1

~22 15.2.4 15.2.7 1 5.2.7 15.2.7 15.2.7 Loss of External Load Turbine Trip Loss of Condenser Vacuum Closure of MSIV Reanalysis reported in Section 6 of this report Bounded by 15.2.1 Bounded by 15.2.1 Bounded by 15.2.1 15.2.6 15.2.9 Loss of Non-emergency AC Power to the Station Auxiliaries Bounded by 15.3.1, 15.2.7 Radiological Consequences bounded by UFSAR D E o

CO

Table 3.1 Summary of St. Lucle Unit 1 Chapter 15 Event Review for this Analysis (cont.)

Cross-Reference Table SRP Event/FSAR Event SRP Event ¹ UFSAR Event ¹ Event Name

~Dis ositioo 15.2.7 15.2.8 Loss of Normal Feedwater Flow Bounded by 15.2.1 for primary system pressurization, bounded by Reference 11 for S. G. Inventory.

1 5.2.8 Feedwater System Pipe Breaks Inside and Outside Containment Bounded by 15.1.5 15.3.1

.1 5.2.5 Loss of Forced Reactor Coolant Flow (4 pump coastdown)

Plant response bounded by Cycle 6 analysis.

Analyzed for DNBR in DNB LCO Verification 15.3.3 15.3.4 Reactor Coolant Pump Rotor Seizure Reanalysis reported in Section 6 of this report 15.3.4 15.2.1 Reactor Coolant Pump Shaft Break Uncontrolled CEA Withdrawal from a Subcritical or Low Power Startup Condition Not Part of Licensing Basis Bounded by 15.4.2 15.4.2 15.2.1 Uncontrolled CEA Withdrawal at Power Fast transients plant response bounded by Cycle 6 Analysis.

Slow CEA withdrawals reanalyzed in Reference 2 15.4.3 15.2.3 CEA Misoperation (Dropped CEA Only)

. Plant Response bounded Cycle 6 Analysis.

Analyzed for DNBR in DNB LCO Verification m

0 K Ql (Q

~

s~ G)

(D Ol

Table 3.1 Summary of St. Lucle Unit 1 Chapter 15 Event Review for this Analysis (Cont.)

Cross-Reference Table SRP Event/FSAR Event SRP Event 4 15.4.4 UFSAR Evsnt tS 15.2.6 Event Name Startup of an Inactive Loop

~Dts osition Tech Specs do not permit at power 15.4.6 15.2.4 CVCS Malfunction that Results In a Bounded by Reference 2 Decrease ln the Boron Concentration in the Reactor Coolant 15.4.7 15.3.3 inadvertent Loading and Operation of a Fuel Assembly ln an Improper Location Bounded by UFSAR 15.4.8 15.4.5 Spectrum of CEA Ejection Accidents Radiological consequences bounded by Reference

12. Energy deposition bounded by Reference 2.

15.5.1 15.5.2 Inadvertent Operation of ECCS CVCS Malfunction that Increases Reactor Coolant Inventory Not ln plant license basis Not in plant license basis

".5.6.1 15.2.12 Inadvertent Opening of a Pressurizer Bounded by 15.3.1 (MDNBR)

Pressure Relief Valve m

CD M CB 0

CJl

Table 3.1 Summary of St. Lucle Unit 1 Chapter 15 Event Review for this Analysis (Cont.)

Cross-Reference Table SRP Event/FSAR Event SRP Event ¹ UFSAR Event ts Event Name

~Ois osition 15.6.2 Radiological consequences of the Failure of Small Uncs Carrying Primary Coolant Outside Containment Not in plant licensing basis 15.6.3 15.4.4 Radiological Consequences of Steam Generator Tube Failure Bounded by Reference 12

i.6.5 15.4.1 15.3.1 Loss of Coolant Accident (Large Break)

Reanalyzed in Reference 3 Loss of Coolant Accident (Small Break)

Reanalyzed in Reference 4 15.7.1 15.7.2 15.3.2 Minor Secondary System Breaks Deleted Deleted UFSAR conclusions unchanged 15.7.3 11.2, 2.4.12 Postulated Radioactive Release Due to Liquid Containing Tank Failures Bounded by Reference 12 15.7.4 15.4.3 Radiological Consequences of Fuel Handling Accidents Bounded by Reference 12 15.7.5 15.4.3 Spectrum of Cask Drop Accidents Bounded by Reference 12

Table 3.1 Summary of St. Lucle Unit 1 Chapter 15 Event Review for this Analysis (Cont.)

Cross-Reference Table SRP Event/FSAR Event SRP Event st UFSAR Event ¹ 15.2.2 Event Name Asymmetric Events:

Loss of External Load Increase in Steam Flow Loss of Normal Feedwater Flow increase in Feedwater Flow

~Dts osition Bounded by Symmetric Events

EMF-92-165 Page 4-1 4.0

SUMMARY

OF OPERATING LIMITS Operating limitsfor the St. Lucie Unit 1 nuclear plant are summarized below. Methods of analysis for determining or verifying the operating limits are detailed in Section 4.3 and Reference 14.

4.1 Reactor Protection S stem The reactor protection system (RPS) is designed to assure that the reactor is operated in a safe and conservative manner.

The input parameters for the RPS are denoted as limiting safety system settings (LSSS). The values or functional representation of the LSSSs are calculated to ensure adherence to the specified acceptable fuel design limits (SAFDLs) during steady state and AOOs. The safe operation of the reactor is also maintained by restricting reactor operation to be in conformance with the limiting conditions for operation (LCOs) which are administratively applied at the reactor plant. The LSSS and LCO parametric values are presented in the following sections.

4.2 S ecNed Acce table Fuel Desi n Limits The SAFDLs are experimentally or analytically based limits on the fuel and cladding which preclude fuel damage.

These limits should not be exceeded during steady-state operation or during AOOs. The SAFDLs are used to establish the reactor setpoints to ensure safe operation of the reactor.

The specmc SAFDLs used to establish the setpoints are:

~

The local power density (LPD) which coincides with fuel centerline melt,

~

The MDNBR corresponding to the accepted criterion which protects against the occurrence of DNB.

The LPD limitfor St. Lucie Unit 1 has been 21 kW/ftin prior cycles and this limitis being retained for Cycle 12.

It is noted that reload fuel for Cycle 11 contains gadolinia-bearing fuel rods which, for given LPD, willoperate with a higher fuel temperature and willconsequently have a lower LPD limit. The neutronics design of the gadolinia-bearing fuel rods is such that the maximum LPD in the gadolinia-bearing fuel rods, with a standard fuel rod at tie. P1 kW/ft limit, will be sufficiently

EMF-92-1 65 Page 4-2 below 21 kW/ftto prevent centerline melt. Therefore, the gadolinia-bearing fuel will not become limiting and the 21 kW/ft design limitwill remain applicable.

The XNB critical heat flux correlation@ was used in the thermal margin analysis with statistical parameters corresponding to an upper 95/95 value of 1.22 which is conservative relative to the 95/95 limitfor XNB, Observance of the limiting conditions for operation will protect against DNB with 95% probability at a 95% confidence level during an AOO.

The XNB correlation was qualified for application to 14x14 fuel in St. Lucie Unit 1 in Reference 15.

4.3 Set oint Anal sis The following sections verify operation of St. Lucie Unit 1 with the current setpoint barns.

The analyses were performed with bounding input parameters to support operation with a peak assembly average burnup of 52,500 MWd/MTU.

4.3,1 Limitin Safe S stem Settln s 4.3.1.1 Local Power Distribution Control Neither the increase in SGTP nor the decrease in Technical Specification PCS flow affects the core power distributions. Therefore, the Cycle 11 analysis<~ is bounding for this analysis.

4.3.1.2 Thermal Mar in Low Pressure The thermal margin/low pressure (TM/LP) trip protects against the occurrence of DNB during steady state operation and for many, but not all, AOOs. This reactor trip system monitors primary system pressure, core inlet temperature, core power and ASI. A reactor trip occurs when primary system pressure falls below the computed limiting core pressure, P~ As with the LPD trip, a statistical setpofnt methodology<

@ was used to verify the adequacy of the existing TM/LP trip for the decreased Technical Specification PCS flow of 355,000 gpm. The TM/LP setpoint was verified using the power distributions of Cycle 11 as a basis.

The methodology for the TM/LP

EMF-92-165 Page 4-3 trip accounts for uncertainties in core operating conditions, XNB DNB correlation uncertainties, and uncertainties in power peaking.

The existing TM/LP trip function for operation at 2700 MWt is given by:

Pvar 2061 XA1(ASI)XQR1(Q)+1 5 85 XTI -8950 where Q is the higher of the thermal power'and the nuclear flux power, T,.n is the inlet temperature in 'F and A1 and QR1 are shown in Figures 4.1 and 4.2, respectively, The uncertainties shown in Tables 4.1 and 4.2 were included in the verification of the TM/LP trip as described in Reference 16, Axial power profiles and scram curves for Cycle 11 were included in this analysis.

A bias (penalty) determined from slow CEA withdrawal calculations, to account for the increase In RTD delay time to 16 sec, was included in determining the TM/LP margin for this analysis.

An excess margin of protection is provided by the existing trip for Cycle 11 including the decreased Technical Specification PCS flow.

4.3.2 Additional Tri Functions, In addition to the TM/LPtrip function, other reactor system trips have been determined to provide adherence to reactor system design criteria. The setpoints for these trips, shown in Table 4.3, are unchanged from the Cycle 6 values except for the SG low level trip setpoint.

4.3.3 Umitin Conditions for 0 eration

~NM It N

The TM/LP trip system does not monitor reactor coolant flow and does not consider changes in power peaking which do not significantly change ASI. Thus, the TM/LP trip generally does not provide DNB protection for the pump coastdown and CEA drop AOOs.

The analysis of these transients is given in References 8 and 17. The LCO presented here administratively protects the DNB SAFDL for these transients,

EMF-92-165 Page 4-4 The method used to establish the DNB LCO involved simulations of the CEA drop and the Loss-of-Flow transients.

The core thermal hydraulic code XCOBRA-IIIC< > is used to determine the initial power, as a function of ASI, which provides protection from DNB with 95% probability. The uncertainties listed in Tables 4.4 and 4.5 were applied using the methodology described in Reference 16.

The validity of the existing DNB LCO for allowable core power as a function of ASI 'was verified to ensure adherence to the SAFDL on DNB during CEA drop and Loss-of-Flow AOOs.

The statistical analysis accounted for the effects of uncertainties associated with incore operating parameters, the XNB critical heat fiux correlation, and power peaking.

Axial power profiles and scram curves for Cycle 11 were included in the analysis. The decreased Technical Specification PCS flow of 355,000 gpm was also included in this analysis.

The allowed core power as a function of ASI is shown in Figure 4.3 and verifies the adequacy of the DNB LCO barn.

4.3.3.2 Unear Heat Rate Monitorln Neither the increase in SGTP nor the decrease in Technical SpeciTication PCS flow affects the core power distributions, Therefore, the Cycle 11 a."alysis<~ is bounding for this analysis.

EMF-92-165 Page 4-5 Table 4.1 Uncertainties Applied in LSSS Calculations Source Engineering tolerance Peaking uncertainty (%)

Trip processing

& decalibration Value-

+ 0.03

+ 8.5

- 0.01 of rated

- 0.05 ASl uncertainty LPD TM/LP

+ 0.06

+ 0.05 The distributions are treated as normal and the uncertainty range represents

+ 2 o

EMF-92-165 Page 4-6 Table 4.2 Uncertainties Applied ln the TNI/LP LSSS Calculations Source Pressure Measurement Trip bias Inlet coolant temperature XNB correlation(@

Flow measurement Value

+ 126.66 psi

-42psi 2'F

• '0*

+ 0.198

+ 0.039 of rated The distributions are treated as normal and the uncertainty range represents

+ 2 o values.

Not treated statisticaliy (determined by Slow CEA Withdrawal event).

A mean value of 1.0571 was used to effectively provide a 95/95 MDNBR limit at 1.22.

EMF-92-1 65 Page 4-7 Table 4.3 Additional LSSS Trip Functions Parameter Low steam gen'erator pressure Low steam generator water level Variable high power Set Point 600 psia 20.5%

9.61% of rated

(<107% of rated)

~Uncerlain 22 psl 4.04%

Low reactor coolant flow High pressurizer pressure Asymmetric steam generator pressure 95%

2400 psia 135 psid 22 psl 22 psl

EMF-92-165 Page 4-8 Table 4.4 Uncertainties Applied in the LCO Calculations Source Engineering tolerance Peaking uncertainty (%)

Power measurement ASI uncertainty Value

+ 0.03

+ 8.5

+ 0.02 of rated

+ 0.06 The distributions are treated as normal and the uncertainty range represents

+ 2 0.

EMF-92-165 Page 4-9 Table 4.5 Uncertainties Applied in DNB LCO Calculations SoUIC8 Pressure measurement Inlet coolant temperature XNB correlation~8)

~CEA Dra

+ 30psi

+ 2.8'F

+ 0.198**

Value-Loss of Flow

+ 22 psl

+ 2'F

+ 0.198**

Flow measurement Scram delay Trip setpoint Scram rod insertion time

'od Worth Flow Coastdown

+ 0.039 of rated

+ 0.039 of rated

+ 0.20 seconds

+ 0.02 of rated flow

+ 0.266 seconds

+ 0.628% h,p

+ 0.098

• seconds The distributions are treated as normal and the uncertainty range represents

+2 o.

A mean value of 1.0571 was used to effectively provide a 95/95 MDNBR limit of 1.22.

Time for flow to drop from 90% to 80% of rated flow.

1.40 l '('

l Oz LLzO I

O QJ OO CL I

CL I

1.35 1.30 1.25 1.20 1.15 1.10 1.05

-0.6

-0.5

-0.4 3

-0.2

-0.1 0.0 0.1 0.2 0.3 0-4 0.5 0.6 AXIALSHAPE INDEX FIGURE 4.1 THERMAL MARGIN/LOWPRESSURE TRIP SETPOINT PART 1

1.2 (1.200,1.200)

P y~ = 2061 +A1+QR1 + 15.85 T< N 8950 (0.972.0.972)

O.B (0.781,0.863) 0.6 0.4 0.2 (0.000,0.235) 0.2 OA 0.6 O.B FRACTION OF RATED THERMAL POWER 1.2 FIGURE 4.2 THERMAL MARGIN/LOWPRESSURE TRIP SETPOINT, PART 2

160 140 0

CEA DROP h

LOSS OF FLOW 0

TECH. SPEC. BARN 0

0 0

R 5

+0 100 CL IJJ 0

80 C5 0

60 yb 4

(-o.5,65)

(-0.08,100)

(0.15,100)

(o.5,65) 40 20

.60

.50

.40

.30

.20'

.10 0

.10

.20

.30

.40

.50

.60 PERIPHERAL AXIALSHAPE INDEX FIGURE 4.3 VERIFICATION OF CYCLE DNB LIMITINGCONDITION OF OPERATION t

(WITH A TECHNICAL SPECIFICATIO)W OF 355.000 GPM}

T

EMF-92-1 65 Page 5-1 5.0 LOSS OF EXTERNALLOAD A Loss of External Load event is initiated by either a loss of external electrical load or a turbine trip. Upon either of these two conditions, the turbine stop valve is assumed to rapidly close Normally, a reactor trip would occur on a turbine trip, However, to calculate a conservative system

response, the reactor trip on turbine trip is disabled.

The steam dump system (atmospheric dump valves-ADVs) is assumed to be unavailable, These assumptions allow the Loss of External Load event to bound the consequences of: Event 15.2.2 (Turbine Trip-steam dump system available);

Event 15.2.3 (Loss of Condenser Vacuum-steam dump system unavailable); and, Event 15.2.4 (MSIV Closure).

The Loss of External Load event primarily challenges the acceptance criteria on primary system overyressurization and DNBR.

The event results in an increase in the primary system temperatures due to an increase in the secondary side temperature.

As the primary system temperatures

increase, the coolant expands into the pressurizer causing an increase in the pressurizer

pressure, The primary system is protected against over-pressurization by the pressurizer safety and relief valves.

Pressure relief on the secondary side is afforded by the steam line safety/relief valves. Actuation of the primary and secondary system safety valves limits the magnitude of the primary system temperature and pressure increase.

With a positive moderator temperature coefficient, increasing primary system temperature results in an increase in core power. The increasing primary side temperatures and power reduce the margin to thermal limits (i.e., DNBR limits) and challenge the DNBR acceptance criteria.

5.2 Definition of Events Ana ed The objectives in analyzing this event are to demonstrate that the primary pressure relief capacity is sufficient to limitthe pressure to less than 110% of the design pressure (2750 psia) and that

EMF-92-165 Page 5-2 the minimum DNBR remains above the safety limit. No.credit is taken for direct reactor trip on turbine trip, the turbine bypass system or the steam dump system.

Two cases are normally analyzed for this event: one challenging overpressurization criteria, and one challenging the fuel design limit(MDNBRCase).

This event was analyzed for Cycle 6 for the first reload of SPC fuel.<@

The analysis showed that the standpoint of DNB, this event is bounded by the loss of coolant flow event.

The increase in SGTP and decreased in Technical Specification PCS flow will tend to reduce the calculated MDNBRs for all of the ONB events in a similar manner.

Therefore, this event willcontinue to be bounded by the loss of coolant flow event.

Reanalysis of the event is not required.

The analysis of this event ts performed at a Technical Specification PCS flow rate of 355,000 gpm. The peak pressurizer pressure is calculated to be 2595 psia.

Using this peak pressurizer pressure, the peak system pressure becomes 2701 psia at the outlet of the PCS pumps.

This is below the criteria of 110% of design pressure (2750 psla), which satisfies the overpressure design criteria.

The peak SG secondary pressure was calculated to be 1022 psia.

This is below the criteria of 110% of design pressure (1100 psla), which satisfied the overpressure design criteria.

Table 5.1 on the following page gives a sequence of events summary forthis analysis.

Following Table 5.1 are a series of figures plotting the following parameters:

Reactor Power, Core Average Heat Flux, Reactor Coolant System Temperature, Pressurizer

Pressure, Reactivity Levels, and Secondary Pressure.

To aid in the reading of the plots, Table 5.2 gives a listing of the

. nomenclatuie used,

EMF-92-1 65 Page 5-3 Table 5.1 Summary of Events for the Loss of External Load EVENT Turbine Trip Pressurizer Heaters on Reactor Scram (Begin Rod insertion) on High Pressurizer Pressure Pressurizer Safety Valve Opens Peak Power Steam LJne Safety Valves Open Peak Primary Pressure Peak Core Average Temperature Peak Steam Dome Pressure TlME (Sec) 0.00 0.00 3.80 4.20 4.27 5.11 5.20 6.78 7.50

EMF-92-1 65 Page 5-4 Table 5.2 Summary of Events for the Loss of External Load VARIABLE NAME DK DKDOP DKMOD TAVG1 TCIO TCL1 THL1 PDO1 PL PPR QOA DEFINITION Total Reactivity Doppler Reactivity Moderator Temperature Reactivity Core Average Temperature, Loop 1 Core Inlet Temperature Cold Leg Temperature, Loop 1 Hot Leg Temperature, Loop 1 Steam Generator Dome Pressure, Loop 1 Core power Level Pressurizer Pressure Core Average Heat Flux

3000 PL 2500 2000 Cr 1500 ELJ O

CL 1000 500 2.5 7.5 10 TIME, sec 12.5 15 17.5 FIGURE 5.1 REACTOR POWER LEVEL FOR LOSS OF EXTERNAL LOAD 20 m

O~

Q)

~

(QI Ul f)

Vl Ol

200000 QOA CV 4

I L

X l~~r

~)

L~ I QJ 175000 150000 125000 100000 75000 50000 0

2.5 7.5 10 TIME, sec 12.5 15 17.5 20 FIGURE 5.2 CORE AVERAGE HEAT FLUX FOR LOSS OF EXTERNAL LOAD

610 600 TAVG1 TCIO

TCL1 THLl 590

~ l~

580 bJ

~ CY 570 CL LLJ 560 550

~I r

r r/r 540 2.5 7.5 10 TIME, sec 12.5 15 17.5 20 FIGURE 5.3 REACTOR COOLANT SYSTEM TEMPERATURES FOR LOSS OF EXTERNAL LOAD

2700 2600 PPR O

'J CL CA V) bJ CL lY.

LLJ M

(f)

Ld 0

2500 2400 2300 2200 2100 2000 2.5 7.5 10 TIME, sec 12.5 15 17.5 20

0 M -2 O

'0 0-4 I

O LLI 5

DK DKDOP DKMOD 0

2.5 7.5 10 TIME, sec 12.5 15 17.5 20 FIGURE 5.5 REACTIVITIES FOR LOSS OF EXTERNAL LOAD

1075 1050 P001 1025 O

1000 LLJ 975 CA V) 950 CL 925 O

Cl 900 875 850 825 2.5

~

7.5 10 TIME, sec 12.5 15 17.5 20 FIGURE 5.6 SECONDARY PRESSURE FOR LOSS OF EXTERNAL LOAD

EMF-92-165 Page 6-1 6.0 REACTOR COOLANT PUMP ROTOR SEIZURE 6.1 This event is initiated by a seizure of a PCS pump rotor.

The seizure causes an immediate reduction fn PCS flow rate, As in the Loss of Forced Coolant Flow event (Event 15.3.1), the impact of losing an RCS pump is a decrease in the active flow rate in the reactor core and, consequently, an increase in core temperatures.

Prior to reactor trip, the combination of decreased flow and increased temperature poses a challenge to DNB limits.

The event is terminated by the PCS low flow trip.

6.2 finition of vents Ana ed One case is analyzed for this event to maximize the challenge to the DNB limit. A new system calculation was not performed for this analysis,

Instead, a set of conservative boundary conditions were assumed utilizing the original system calculation as a basis, A Technical SpecNcatlon PCS flowof 355,000 gpm with an uncertainty of +/-14,500 gpm is used in this analysis.

A calculation of the percent of fuel rods in the core to experience DNB as a result of the event is performed. A statistical treatment of the pressure, the inlet temperature, the PCS flow, the linear heat generation rate, and the radial peaking factor uncertainties yields a percentage ofrods predicted to experience DNB. The calculation employed the XNBcritical heat flux correlation@ with a 95/95 DNBR limit of 1.22.

The percentage of rods predicted to experience DNB was 1,00%.

The licensing basis radiological consequences analyslsi' assumed 2.5% of the rods in the core failed, This bounds the failures calculated in the current analysis, Therefore, the results of the reference analysis bound operation with reduced PCS flow rate.

EMF-92-165 Page 6-2 Radiological consequences for the reference analysis are bounded by the LOCA event, as described in Reference 12.

EMF-92-165 Page 7-1

7.0 REFERENCES

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~

St.

ucf Unit 1 Lar Break LOC ECCS Ana sis with 25% Steam Generator Tube Plu in and an As mmet of +- 7%, EMF-92-176, Siemens Power Corporation-Nuclear Division, Richland WA 99352, February 1993.

St, Lucie Unit 1 Small LOCA Break Anal sis, EMF-92-148, Siemens Power Corporation-Nuclear Division, Richland WA 99352, February 1993, XCO RA-IIIC:

A Com uter Code to Determine the Distribution of Coolant Durin Stea state and Transient Core 0 eratlon, XN-NFL-12, Exxon Nuclear Company, Richland, WA 99352, January 1986.

XNBDNB Correlation for PWR Fuel Desi ns, XN-NF421 (P)(A), Revision 1, Exxon Nuclear Company, Richland, WA 99352, September 1982.

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Plan Transient Ana sis for St. Lucie Unit 1 Reactor, XN-NFL-99, Exxon Nuclear Company, Richland, WA 99352, January 1983.

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e Unit Losswf-Feedwater Transient with Reduced Steam enerator Low Level Tri getgg~lf, ANF49-113, Advanced Nuclear Fuel Corporation,

Richland, WA 99352, September 1989.

St. Lucie Unit 1 Assessment of Radiolo ical and Rod Bow Effects for Increased Burnu, ANF48-113(P), Advanced Nuclear Fuels Corporation, Richland, WA 99352, July 1988.

A Generic Ana sis of the Control Rod E ection Transient for Pressurized Water Reactors, XN-NF-78-44(A), Exxon Nuclear Company, Richland, WA 99352, January 1979,

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LOCA ECCS Extended Ex osure Anal sis, ANF-87-148, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1987.

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EMF-92-1 65 Issue Date:

2 9 93 St. Lucle Unit 1 Chapter 15 Event Review and Analysis for 25% Steam Generator Tube Plugging DISTRIBuTION RA Copeland CY Chou KM Ouggan RC Gottula JS Holm SE Spangler BD Stitt Rl Wescott Document Control FPL (11)/JL Holm

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