ML17228B505

Chapter 15 Event Review & Analysis for 30% SG Tube Plugging.

ML17228B505
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Site:	Saint Lucie
Issue date:	05/31/1996
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EMF-96-135, NUDOCS 9606030193
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EMF-96-1 35 St. Lucie Unit 'l Chapter 15 Event Review and Analysis for 30% Steam Generator Tube Plugging May 1996 Siemens Power Corporation Nuclear Division 9606030i93 96060i 05000335 PDR ADDCK P PDR

0 Siemens Power Corporation - Nuclear Division EMF-96-1 35 Issue Date: 5/3g/g6 St. Lucie Unit 1 Chapter 15 Event Review and Analysis for 30% Steam Generator Tube Plugging Prepared by:

B. D. S t, Senior Engineer PWR Safety Analysis Nuclear Engineering May 1996

e 0

EMF-96-1 35 Page i Table of Contents Section Pacae

1.0 INTRODUCTION

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

210

SUMMARY

e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3.0 CHAPTER 15 EVENT REVIEW 3.1 Changes in Technical Specifications and Plant Configuration ..

~ 3 1 3-1 3.2 Changes in Core Physics Parameters ..... ~ . ~ ~.......... 3-1 3.3 Review of Events 3-2 3.2.1 Increase in the Heat Removal by the Secondary System 0 3-2 3.2.2 Decrease In Heat Removal By The Secondary System . ~ 3-4 3.2.3 Decrease in Reactor Coolant System Flow Rate 3-9 3.2.4 Reactivity and Power Distribution Anomalies 3-10 3.2.5 Increase in Reactor Coolant Inventory 3.2,6 Decrease in Reactor Coolant Inventory

....,........ 3-14 3-14 3.2,7 Radioactive Releases from a System or Component 3-18 3.2,8 Asymmetric Events 3-18 4.0

SUMMARY

OF OPERATING LIMITS . 4-1

~

4.1 LPD LSSS and LPD LCO 4-1 4.2 TM/LP 4-1 4.3 DNB LCO 4-1 5.0 LOSS OF EXTERNAL LOAD 5-1 5.1 Event Description ~ 5-1 5.2 Definition of Events Analyzed 5-1 5.3 Analysis Results ~ 5-2 6.0 REVIEW OF MECHANICALDESIGN ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

7.0 REFERENCES

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

APPENDIX A ........ ~..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

EMF-96-1 35 Page ii List of Tables Table Pacae 3.1 Summary of St. Lucie Unit 1 Chapter 15 Event Review for Cycle 14 ...... 3-20 5.1'ummary of Events for the Loss of External Load.............. ~...... 5-3

J EMF-96-1 35 Page iii List of Fi ures

~Fi ure Pacae 5.1 Reactor Power Level for Loss of External Load .... ~ .. ~............... 5-4 5.2 Core Average Heat Flux for Loss of External Load ..............,..... 5-5 5.3 Reactor Coolant System Temperatures for Loss of External Load ... ~.....

~ 5-6 5.4 Pressurizer Pressure for Loss of External Load .......-;................ 5-7 5.5 Reactivities for Loss of External Load ............................. 5-8 5.6 Secondary Pressure for Loss of External Load ....................... 5-9

EMF-96-1 35 Page 1-1 St. Lucie Unit 1 Chapter 15 Event Review and Analysis for 30% Steam Generator Tube Plugging

1.0 INTRODUCTION

This analysis supports a Steam Generator Tube Plugging (SGTP) level of 30% with an asymmetry of +/-7% and a minimum Technical Specification limit on Primary Coolant System (PCS) flow of 345,000 gpm.i'I The analysis also supports a Reactor Coolant Low Flow trip setpoint of 93%3% of design flow (345,000 gpm), " The remainder of the conditions used for this analysis are based on those reported for St. Lucie Unit 1 Cycle 14.( I The topics addressed in this report include a review of the Standard Review Plant (SRP) Chapter 15 Events, Summary of Operating Limits (Setpoints), Loss of External Load analysis (PCS pressurization event) (15.2.1), and results of the mechanical design review.

EMF-96-1 35 Page 2-1 2.0

SUMMARY

A review of SRP Chapter 15 events was performed to assess the impact of an increase in

~

SGTP to 30%, with an asymmetry of +/- 7%, a reduced minimum Technical Specification PCS flow of 345,000 gpm, and a reduced Reactor Coolant Low Flow trip setpoint of 93 a 3%

of design flow (345,000 gpm). The events identified that require reanalysis are listed below.

All other events are either bounded by another event or are bounded by existing analyses of record. This review uses the Cycle 14 event review as a basis.

EVENT SECTION REQUIRING OF ANALYSIS EVENT DESCRIPTION REPORT SECTION DESCRIPTION 1 5.2.1 Loss of External Load 5.0 Loss of External Load 1 5.2.7 Loss of Normal Feedwater 3.2.2.3 Loss of Normal Feedwater 1 5.3.1 Loss of Forced Reactor 4.3 Setpoint Analysis Coolant Flow 1 5.4.3 CEA Misoperation (Dropped 4.3 Setpoint Analysis CEA Only) 1 5.4.6 Decrease of Boron 3.2.4.5 Decrease of Boron Concentration Concentration 1 5.6.5 Small Break LOCA 3.2.6.4 Loss of Coolant Analysis (Note 1)

Note 1: The current licensing basis small break LOCA analysis bounds plant operation up to a Cycle 14 cycle average burnup of 9135 MWd/MTU. Operation at

.burnups greater than this will either require a reanalysis of the SBLOCA event, or a decrease in allowed reactor power to 90% of design power.

To insure Departure from Nucleate Boiling (DNB) criteria are still met with the increased SGTP, the decreased Technical Specification PCS flow, and the decreased Reactor Coolant Low Flow trip setpoint, the Thermal Margin / Low Pressure (TM/LP) and DNB Limiting Condition of Operation (LCO) setpoint analyses were reanalyzed in this report (Section 4). The TM/LP trip setpoint analysis shows an excess margin of protection is provided by the existing trip

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EMF-96-1 35 Page 2-2 equation. The validity of the existing DNB LCO barn for allowable core power as a function of Axial Shape Index (ASI) was verified to ensure adherence to the Specified Acceptable Fuel Design Limits (SAFDL) on DNB during CEA drop and Loss of Flow Anticipated Operational Occurrences (AOOs).

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

The Loss of Normal Feedwater event reanalyses demonstrate that adequate steam generator inventory remains bound to steam generator dryout criteria (Section 3.2.2.3).

The Boron Dilution event reanalyses demonstrate that the calculated times to lose the required shutdown margin are greater than acceptance criteria for all modes of operation (Section 3.2.4.5).

The fuel mechanical design has been evaluated for the reduced flow operating condition, The results confirm that mechanical design parameters will remain essentially unchanged with no significant degradation of any design margins.

EMF-96-1 35 Page 3-1 3.0 CHAPTER 15 EVENT REVIEW A review of the Chapter 15 events for St. Lucie Unit 1 was performed to support an increase in SGTP to 30% with an asymmetry of +I- 7%, and a Technical Specification PCS flow of 345,000 gpm.I~I The review also considered a Reactor Coolant Low Flow trip setpoint of 93 a3% of design flow.I~>> The remainder of the conditions assumed in this analysis are the same as those used in the Cycle 14 safety analyses. The review also considered changes in core physics parameters caused by these changes. There are no changes in the fuel design, I

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, The effect of the changes in core physics parameters is described in Section 3.2. The review of the Chapter 15 events is given in Section 3.3 3.1 Chan es in Technical S ecifications and Plant Confi uration Following is a list of the changes considered in this event review:

Original This Parameter ~Cele 14 Review iii Steam Generator Tube Plugging 25% 30%

Technical Specification 355,000 gpm 345,000 gpm PCS Flow Rate Limit Reactor Coolant Low Flow Trip Setpoint 95% 93%

3.2 Chan es in Core Ph sics Parameters A review of the changes in core physics parameters caused by the changes-described in Section 3.1 was performed. This review concluded that all physics parameters were within bounds used in supporting analyses.

EMF-96-1 35 Page 3-2 3.3 Review of Events The events are presented in the same order as that used in the SRP. A cross reference of SRP and UFSAR 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 for the review. Event acceptance and single failure criteria are those presented in the FSAR for each event. No change in event initiators was considered in the review process.

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, The Reactor Coolant Low Flow trip is only credited in two events, the Loss of Forced Reactor Coolant Flow (15.3.1), and the Reactor Coolant Pump Rotor Seizure (15.3.3). The impact of the reduction in the low flow trip setpoint will be discussed only for these events, A summary of the results of the event review is presented in Table 3.1.

3.2.1 Increase in the Heat Removal b the Secondar S stem 15.1 The events in this category of events were evaluated by calculating the increase in system

'ooling due to the event initiator. Reference 2 dispositioned the AOOs in this category to be bounded from the standpoint of DNB ratio (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 Event (15.1,3) remains the bounding event.

Only Event 15 1.3 is discussed relative to DNBR.

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EMF-96-1 35 Page 3-3 3.2.1.1

~ ~ ~ Increase in Steam Flow 15.1.3 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 for this event is similar to the feedwater malfunction events. The steam demand increase results in depressurization 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.

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 Coolant Flow event, The increase in SGTP and decrease in Technical Specification PCS flow will tend to offset the calculated Minimu'm DNBRs (MDNBR) for all the DNB events in a similar manner. Therefore, this event will continue to be bounded by the Loss of Coolant Flow event. Reanalysis of the event is not required.

Subevent 3: The UFSAR (Table 15.2.11-4) shows that the reactor trips on the Variable High Power Trip (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.

EMF-96-1 35 Page 3-4

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3.2.1.2 Inadvertent 0 enin of

~ ~ ~ a Steam Generator Relief or Safet 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 14.

3.2.1.3 Steam S stem Pi in 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 Maim Steam Isolation Valves (MSIVs) stops blowdown of the intact steam generator a few seconds after event initiation. Secondary blowdown and depressurization cause a cooldown of the moderator, resulting in an erosion of shutdown margin and possible return to power.

The event is most limiting at End-of-Cycle (EOC) conditions, The event was analyzed in Reference 4 for Cycle 6 and on a confirmatory basis using advanced analytical methods in Reference 7. The referenced analysis(7I conservatively assumes no SGTP. Increased tube plugging would reduce the cooldown rate of the PCS and subsequent return 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.

Radiologic'al consequences of this event are bounded by that of the Loss of Coolant Accident (LOCA), event 15.6.5.

3.2.2 Decrease In Heat Removal B The Secondar 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

e EMF-96-1 35 Page 3-5 primary coolant system. At the beginning of cycle when the Moderator Temperature Coefficient (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 15.3.1) is initiated with turbine stop valve closure, a fast acting (0.1 to 0.3 sec. closure) valve.

~ 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 (TT) (15.2.2), Loss of Condenser Vacuum (15.2.3), and Closure of Main Steam Isolation Valve (MSIV) (15.2.4). The use of the

'I 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 MSIV closure time is greater than turbine stop valve closure time. The more rapid closure of the stop valves produces a more severe system transient than does 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 Loss of External Load 15.2.1 I

There are two subevents in the LOEL event; PCS pressurization and DNB. This event was analyzed for Cycle 6 for the first reload of SPC fuel. I The analysis showed that from the

e EMF-96-135 Page 3-6 standpoint of DNB, this event is bounded by the Loss of Coolant Flow event. The increase in SGTP does not affect DNBR. The 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 Loss 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 and the safety valves open. The increased SGTP causes initial steam generator 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 primary safety valves. Both of these effects will tend to produce a higher peak PCS pressure.

Therefore, this event is reanalyzed for increased SGTP for the peak pressure event in 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 limits which remain unchanged from those used in the UFSAR analysis. Therefore, the UFSAR analysis remains bounding.

3.2.2.2 Loss of Non-emer enc AC Power to the Station Auxiliaries 15.2.6 The loss of AC power to the station auxiliaries event results in:

~ Loss of Primary Coolant Pumps

~ Loss of Normal Feedwater

~ Auto Diesel Generator Start

~ Release-of Control Element Drive Mechanism (CEDM) Holding Coils

0 EMF-96-1 35 Page 3-7

~ 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 V

second into the event.

The early part of the event (0-10 seconds) would be the same as the four-pump coastdown event (15.3.1) because the steam generator inventory would not have been reduced sufficiently to affect heat removal. Therefore, the DNB SAFDL and over-pressurization limit are 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 steam generator liquid level trip. Therefore, this event is 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.

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 steam generator 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 steam generator 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.

EMF-96-1 35 Page 3-8 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 relative to overpressurization.

The Loss-of-Feedwater event was reanalyzed relative to steam generator inventory in Reference 8. 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 Specification PCS flow rate will have a small effect on the minimum steam generator inventory event. A reduced flow rate would result in a slightly higher average primary coolant temperature, which increases the energy stored in the primary coolant that has to be removed by boiling off the steam generator inventory. ~

The limiting case from the standpoint of secondary inventory is the case with no auxiliary feedwater from Reference 8. This case is analyzed to ensure that the operators have at least 600 seconds to confirm initiation of the auxiliary feedwater system, For this case, an additional 637 Ibm of inventory would be boiled off from the secondary side of the steam generator, reducing the dryout time from 650 seconds to 611,7 seconds. Thus, the steam generators will not dryout before the maximum time delay on auxiliary feedwater of 600 seconds stated in the St. Lucie Unit 1 Technical Specification Bases.

3.2.2.4 Feedwater S stem Pi e Breaks Inside and Outside Containment 15.2.8 This event is a cooldown 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 'r'elief rate which produces a more benign event.

EMF-96-1 35 Page 3-9 The radiological consequences of this event are bounded by the LOCA event (15.6.5).

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 current analysis was performed with a Technical Specification PCS flow rate of 355,000 gpm. The reduction in Technical Specification PCS flow rate to 345,000 gpm, a 2.8%

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 to be performed. The event was analyzed for the increased SGTP, the reduced Technical Specification PCS flow, and the reduced Reactor Coolant Low Flow trip setpoint by verifying the DNB LCO barn (Section 4.3).

3.2.3.2 Reactor Coolant Pum Rotor Seizure 15.3.3 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 subevent or a pressure subevent.

Pressure Sub-event The event was evaluated in Reference 5 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 Specification PCS flow rate will significantly affect the outcome of this event.

EMF-96-1 35 Page 3-10 DNB Sub-event The same arguments regarding the rate of flow coastdown with increased SGTP and a decrease in the Technical Specification PCS flow rate for event 15.3.1 (Loss of Reactor Coolant Flow) apply for this event. However, a decrease in the Technical Specification PCS flow rate and the reduced Reactor Coolant Low Flow trip setpoint will result in lower DNBRs and a higher percentage of fuel rod failures for this event. Excess conservatism in the reactor power and F, used in the current licensing basis analysis nearly offsets the effects of the decreased coolant flow and-decreased low flow trip setpoint, resulting in a net power penalty of 0.57%. The current licensing basis analysis resulted in 1% fuel failures. However, the radiological consequences analysis for the event assumed 2.5% fuel failures.( ) The small decrease in DNBR associated with the 0.57% power penalty will not cause the number of fuel failures to exceed the 2.5% value used in the radiological analysis, particularly with the" statistical treatment of power used in the seized rotor analysis. Therefore, the current radiological analysis of record bounds operation with the decreased coolant flow and decreased low flow trip setpoint.

3.2.3.3 Reactor Coolant Pum Shaft Break The event is not in the UFSAR, and therefore, not part of the licensing basis.

W 3.2.4 Reactivit and Power Distribution Anomalies 3.2.4.1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Startu Condition 1 5.4.1 The event is analyzed in Section 15.2.2 of the St. Lucie Unit 1 UFSAR. The event was initiated from hot zero power condition by rod withdrawal from a very low initial power level.

Significant 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 terminated by the VHPT. The peak neutron power was high (150% rated), but the time from trip to MDNBR was short (2 seconds), resulting in relatively low surface heat flux and consequently, a

EMF-96-1 35 Page 3-11 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 4. This event is bounded with respect to MDNBR by the Loss of 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, increased SGTP, and decreased Reactor Coolant Low Flow trip setpoint in the DNB LCO analysis, Section 4.3, Therefore, no reanalysis of the Fast CEA Withdrawal event is required.

Slower CEA withdrawals may terminate on the TM/LP trip. The TM/LP setpoint provides designed protection of DNBR limits. The slow URW at hot full power is used to set the pressure bias term used in the TM/LP analysis. The reduced loop flow rate would increase loop transit time, and increase the time delay on the TM/LP trip. A review of this calculation concluded that the change in the loop transit time would have an insignificant effect on the calculated pressure bias term. Therefore,. reanalysis is not required.

\

3.2 4.3 CEA Miso eration 15.4.3 The SRP defines four sub-events:

1) Static Misalignment of CEAs

2) Single CEA Withdrawal

3) Dropped CEA Bank

4) Dropped CEA

EMF-96-1 35 Page 3-12 The first three subevents are not reported in the UFSAR or Reference 4, 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 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 determined by r

the moderator and Doppler feedback. The PCS pressure is determined by the average coolant temperature. Because bounding neutron kinetics parameters are unchanged for Cycle 14 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 verification of the DNB LCO setpoint (Section 4.3). 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 15.4.4 Part loop operation is not permitted by the plant Technical Specification. Therefore, analysis of this event is not necessary.

3.2.4.5 CVCS Malfunctions that Result in a Decrease In the Boron Concentration in the Reactor Coolant 15.4.6 The event can occur during all modes of operation. The modes considered in the UFSAR and Cycle 6("I analyses for this event are; Power (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Cold Shutdown (Mode 5) r Refueling (Mode 6)

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EMF-96-1 35 Page 3-13 The SRP acceptance criteria require protecting the fuel SAFDLs for the at power subevent.

The FSAR presents a verbal analysis which concludes that the TM/LP trip, Local Power Density Limiting Safety System Setpoint (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 14 and is also retained for the increased SGTP level and corresponding decrease in Technical Specification PCS flow rate.

The acceptance criterion for Modes 2 through 6 is that the time to critically allows the operator to terminate the event. The time to criticality was reanalyzed for Modes 2 through 6 for Cycle 14 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 increase in SGTP from 25% to 30% results in a small reduction in PCS inventory of 1.28%. This small reduction in PCS inventory results in a small reduction in the time to lose the required shutdown margin for Modes 2 through 5. The time to criticality was reduced from 72.02 minutes to 71.1 minutes for Modes 2, 3, and 4, relative to the criteria of 15 minutes. The time to criticality was reduced from 20.54 minutes to 20.3 minutes for Mode 5, relative to the criteria of 15 minutes. The effect 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 eration of a Fuel Assembl In An Im ro er The UFSAR presents two subevents:

~ Misloaded Fuel Pellets or Fuel Rod(s) 0

~ Erroneous Placement or Orientation of Fuel Assemblies The discussion presented in the UFSAR is applicable to operation with the increased SGTP and the decreased Technical Specification PCS flow.

0 EMF-96-1 35 Page 3-14 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 of this 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 14 in Reference 2. Increased SGTP has no significant effect 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 flow will also not affect this event. No reanalysis is necessary.

3.2.5 Increase in Reactor Coolant Inventor 3.2.5.1 Inadvertent 0 eration of ECCS 15.5.1 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 Inventor 15.5.2 The event is not in the UFSAR and therefore not part of the licensing basis for the plant.

3.2.6 Decrease in Reactor Coolant Inventor 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 7 as a PCS depressurization event. The event is bounded with respect to DNBR by the Loss-of-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

EMF-96-1 35 Page 3-15 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 Carr in Primar Coolant Outside Containment 15.6.4 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 steam generator tube is choked before reactor scram. Therefore, the reduction in secondary pressure due to increased SGTP during this time doesn't 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 steam generator 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 6 addresses the radiological consequences for increased burnup fuel. The decrease in Technical Specification PCS flow does not affect this event. No reanalysis is required.

3.2.6.4 Loss of Coolant Accident LOCA 15.6.5 Lar e Break Increased SGTP from 25% to 30% and decreased PCS flow from 355,000 to 345,000 gpm will cause an increase in PCS average coolant temperature of 0.77 F and a 6 psi decrease in the secondary pressure during normal operation. These small changes in operating conditions will have small offsetting effects during the blowdown and ref lood portions of the Large Break LOCA (LBLOCA) event. Based on SPC methodology and the results of the current licensing analysis, these changes will have essentially no effect on the calculated peak cladding

EMF-96-1 35 Page 3-16 temperature (PCT) during the ref lood portion of the transient. There will be minor changes in the blowdown characteristics which may or may not cause a small increase in the PCT.

There are conservatisms in the licensing analysis that would dominate and offset any small increase in PCT that might be caused by increased SGTP and decreased PCS flow. Two significant parameters in the LBLOCA analysis are initial fuel stored energy and the elevation in the core where the peak power occurs.

The LBLOCA licensing analysis is performed in two parts which conservatively combine initial fuel stored energy and axial power shapes. The first part uses an initial fuel stored energy near BOC where maximum densification occurs, combined with a bounding shape at Middle-of-Cycle (MOC) peaked at a relative core height of 0.77. The limiting MOC axial shape for Cycle 14 is peaked at a relative core height of 0.65. The difference in elevation of peak power at MOC for Cycle 14 relative to that used in the licensing analysis represents significant margin in PCT during the ref lood portion of the transient. In addition, a maximum resinter density of 1.1% is used in the licensing analysis compared to the as-built resinter density for Cycle 14 of 0.81%. Use of an as-built resinter density would result in a reduction in initial fuel average temperature of 34'F at the time of maximum densification. This reduction in initial fuel stored energy represents a significant reduction in PCT. Combined, these two conservatisms would overwhelm any small increase in PCT caused by increased SGTP and decreased PCS flow.

The second part of the LBLOCA analysis uses a fuel stored energy representing MOC, combined with a bounding EOC axial poyver shape. Use of an as-built resinter density would result in at least a 16'F reduction in fuel average temperature compared to that used in the licensing analysis. In addition, a conservatively low burnup representing MOC was used in the licensing analysis relative to a MOC burnup for Cycle 14. Since the fuel average temperature reduces significantly from the time of maximum densification to the EOC, the fuel average temperature used in the licensing analysis at MOC is conservative by about 90 F relative to Cycle 14. This amount of conservatism in fuel average temperature represents a very large conservatism in PCT. Combined, these two conservatisms would overwhelm any small increase in PCT caused by increased SGTP and decreased PCS flow.

EMF-96-1 35 Page 3-17 Therefore, the LBLOCA analysis of record continues to bound operation at full power with increased SGTP and decreased PCS flow rate.

Small Break An increase in SGTP from 25% to 30% causes a small reduction in the initial PCS coolant inventory, which causes a slight decrease in core liquid level and increase in PCT. Also, the decreased PCS flow rate causes a small increase in the primary coolant average temperature, which tends to hold primary system pressure higher early in the event and increase the mass lost out of the break. However, these effects are relatively small. The analysis uses a top-peaked axial profile, which is obtained from EOC conditions. The peak power location is high in the core, which is limiting because it stays uncovered for the longest period of time, and also deposits the greatest amount of energy above the liquid mixture level, which maximizes the steam temperatures and minimizes the production of steam. The effect of the elevation of the peak power point in the core greatly outweighs the smaller effects of the increased SGTP and reduced loop flow.

The elevation of the peak power point will steadily increase with burnup. Therefore, the effect of using a bounding EOC axial profile will offset the effects of increased SGTP and decreased loop flow for burnups at least to the point where the MOC axial power profiles were generated because the MOC axial power profiles have peak power elevations significantly lower in the core than the EOC axial power profiles. For burnups greater than MOC, it is not possible without a reanalysis to determine when the effects of the increased SGTP and decreased flows would'overshadow the increasing peak power elevation. Therefore, the current Small Break LOCA (SBLOCA) analysis, Reference 11, will bound operation only up to the burnup at which the MOC axial power profiles were generated (9135 MWd/MTU Cycle 14 cycle average burnup).

The current SBLOCA analysis bounds operation up to a Cycle 14 cycle average burnup of 9135 MWd/MTU. To operate at higher burnups will require a re-analysis or a derating of reactor power to 90% of design power. The SBLOCA event is very sensitive to total core power because this is what determines the decay heat which must be removed by boiling off

0 EMF-96-1 35 Page 3-18 the primary system liquid inventory. Reducing the decay heat by 10'/o has a strong impact on the results because the difference in power integrates over the entire course of the event.

This significantly reduces the liquid inventory lost out of the break and results in less core level uncovery and lower PCTs. This reduction in the liquid inventory lost out of the break is twice as large as the reduction in initial inventory caused by the increase in 4'pproximately SGTP from 25'/o to 30'/o. Also, primary system pressure will decrease more rapidly because of the lower steam generation rates at 90/o power. Operation at 90'/o of rated power will also cause a decrease in the primary coolant average temperature which is larger than the increase in temperatures caused by the decrease in PCS flow rate. Therefore, operation at 90'/o of rated power bounds the effects of increased SGTP and decreased PCS flow rate. In order to operate at a power level greater than 90/o of rated power, the SBLOCA event would have to be reperformed to include the increased SGTP and reduced PCS flow rates.

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 7 for extended burnup levels are bounding for an increase in SGTP level and decrease in Technical Specification PCS flow.

3.2.8 As mmetric Events 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.

EMF-96-1 35 Page 3-19 This disposition assumes an increased asymmetric steam generator tube plugging of 37/0 in one steam generator an 23'/0 in the other SG. 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.

e EMF-96-1 35 Page 3-20 Table 3.1 Summary of St. Lucie Unit 1 Chapter 15 Event Review for Cycle 14 SRP/USFSAR Cross-Reference SRP UFSAR Event Name Disposition 15.1.1 15.2.10 Decrease in Feedwater Bounded by 15.1.3 Temperature 1 5.1.2 15.2.10 Increase in Feedwater Flow Bounded by 15.1.3 1 5.1.3 1 5.2.1 1 Increase in Steam Flow Bounded by 15.3.1 15.1 4 15.2.11 Inadvertent Opening of a Bounded by 15.,1.3 Steam Generator Relief or Radiological Consequences Safety Valve bounded by UFSAR 1 5.1.5 1 5 4.6 Steam System Piping Failures Bounded References 4 and 9 Inside and Outside of Containment 15.2.1 1 5.2.7 Loss of External Load Reanalysis Reported In Section 5.0 of this Report 1 5.2.2 1 5.2.7 Turbine Trip Bounded by 15.2.1 1 5.2.3 1 5.2.7 Loss of Condenser Vacuum Bounded by 15.2.1 5.2.4 - 1 5.2.7 1 Closure of MSIV Bounded by 15.2.1 1 5.2.6 1 5.2.9 Loss of Non-emergency AC Bounded by 15.2.1, 15.3.1 Power to the Station Radiological Consequences Auxiliaries bounded by UFSAR 1 5.2.7 1 5.2.8 Loss of Normal Feedwater Bounded by 15.2.1 for primary Flow system pressurization. Limiting case evaluated in Section 3.2.2.3 for Steam Generator Inventory 1 5.2.8 1 5.2.8 Feedwater System Pipe Bounded by 15.1.5 Breaks Inside and Outside Containment 1 5.3.1 1 5.2.5 Loss of Forced Reactor Plant response based on Cycle 6 Coolant Flow (4 pump analysis. Analyzed for DNBR in coastdown) DNB LCO Verification 1 5.3.3 1 5.3.4 Reactor Coolant Pump Rotor Bounded by References 6 and 9 Seizure

EMF-96-1 35 Page 3-21 Table 3.1 Summary of St. Lucie Unit 1 Chapter 15 Event Review for Cycle 14 (Continued)

SRP/USFSAR Cross-Reference SRP UFSAR Event Name 'Disposition 1 5.3.4 Reactor Coolant Pump Shaft Not Part of Licensing Basis Break 1 5.4.1 15.2.1 Uncontrolled CEA Withdrawal Bounded by 15.4.2 from a Subcritical or Low Power Startup Condition 1 5.4.2 15.2.1 Uncontrolled CEA Withdrawal Fast transients plant response at Power bounded by 15.3.1. Slow CEA withdrawals bounded by Reference 2.

1 5.4.3 1 5.2.3 CEA Misoperation (Dropped Plant Response bounded by CEA Only) Cycle 6 Analysis. Analyzed for DNBR in DNB LCO Verification 1 5.4.4 1 5.2.6 Startup of an Inactive Loop Part loop operation not allowed by Technical Specifications 1 5.4.6 15.2.4 CVCS Malfunction that Reanalyzed in Section 3.2A.5 Results in a Decrease in the Boron Concentration in the Reactor Coolant 1 5.4.7 1 5.3.3 Inadvertent Loading and Bounded by UFSAR Operation of a Fuel Assembly in an Improper Location 1 5.4.8 1 5.4.5 Spectrum of CEA Ejection Radiological consequences Accidents bounded by Reference 6.

Energy deposition bounded by Reference 2.

1 5.5.1 Inadvertent Operation of Not Part of Licensing Basis ECCS 1 5.5.2 CVCS Malfunction that Not Part of Licensing Basis Increases Reactor Coolant Inventory 15.6.1 1 5.2.1 2 Inadvertent Opening of a Bounded by 15.3.1 Pressurizer Pressure Relief Valve

EMF-96-1 35 Page 3-22 Table 3.1 Summary of St. Lucie Unit 1 Chapter 15 Event Review for Cycle 14 (Continued)

SRP/USFSAR Cross-Reference SRP UFSAR Event Name Disposition 1 5.6.2 Radiological consequences of Not Part of Licensing Basis the Failure of Small Lines Carrying Primary Coolant Outside Containment I 5.6.3 1 5.4.4 Radiological Consequences of Bounded by Reference 6 Steam Generator Tube Failure 1 5.6.5 1 5.4.1 Loss of Coolant Accident Bounded by Reference 11 up to (Small Break) a cycle average burnup of 9135 Mwd/MTU. Requires power derate to 90% of rated power or reanalysis to operate at higher burnups.

1 5.3.1 Loss of Coolant Accident Bounded by Reference 11 (Large Break) 1 5.3.2 Minor Secondary System UFSAR conclusions unchanged Breaks 15.7.1 15.4.2 Waste Gas System Failure Not,Part of Licensing Basis 15.7.2 Radioactive Liquid Water Not Part of Licensing Basis System Leak or Failure 1 5.7.3 Postulated Radioactive Bounded by Reference 6 Release Due to Liquid Containing Tank Failures 1 5.7.4 1 5.4.3 Radiological Consequences of Bounded by Reference 6 Fuel Handling Accidents 15.7.5 15 4.3 Spectrum of Cask Drop Bounded by Reference 6 Accidents 1 5.2.2 Asymmetric Events: Bounded by Symmetric Events Loss of External Load, Increase in Steam Flow, Loss of Normal Feedwater Flow, Increase in Feedwater Flow

EMF-96-1 35 Page 4-1 4.0

SUMMARY

OF OPERATING LIMITS Operating limits for the St. Lucie Unit 1 nuclear plant are summarized below. The impact of the changes in SGTP, PCS flow, and Reactor Coolant Low Flow trip setpoint are demonstrated in the setpoint analyses. The setpoint analyses include the LPD LSSS, LPD LCO, TM/LP trip, and the DNB LCO.

4.1 LPD LSSS and LPD LCO The LPD LSSS and LPD LCO analyses assure that maximum Linear Heat Generation Rate (LHGR) is below limits to protect fuel centerline melt and Technical Specification LHGR limits, respectively. Neither the increase in SGTP nor the decrease in Technical Specification PCS flow affects the core power distributions. Therefore, the Cycle 14 analysis is bounding for these analyses.

4.2 ~TM LP The TM/LP trip protects against the occurrence of DNB during steady state operation and for many, but not all, AOOs. The TM/LP verification was reanalyzed to include the effects of the reduced PCS flow rate. Axial power profiles and scram curves for Cycle 14 were included in this analysis. An excess margin of protection is provided by the existing trip for Cycle 14 including the decreased Technical Specification PCS flow.

4.3 DNB LCO 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 Loss-of-Flow and CEA drop AOOs. The LCO presented here administratively protects the DNB SAFDL for these transients.

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.

EMF-96-1 35 Page 4-2 The statistical analysis accounted for the effects of uncertainties associated with incore operating parameters, the XNB critical heat flux correlation, and power peaking. Axial power profiles and scram curves for Cycle 14 were included in the analysis. The impact of the decreased Technical Specification PCS flow and the reduction in the Reactor Coolant Low Flow trip were deterministically evaluated to conservatively bound the effects of the changes.

The evaluation resulted in a decrease in the minimum power margin from 8.0 'o 4.6/o of rated power for the CEA drop event, and a decrease from 6.8'lo to 1.9'lo of rated power for the Loss of Flow event. An excess margin of protection is provided by the existing barn for Cycle 14.

EMF-96-1 35 Page 5-1

/

5.0 LOSS OF EXTERNAL LOAD 5.1 Event Descri tion 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 over-pressurization 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 Anal zed The objectives in analyzing this event are to demonstrate that the primary pressure relief capacity is sufficient to limit the pressure to less than 110% of the design pressure (2750 psia) and that the minimum DNBR remains above the safety limit. No credit is taken for direct

EMF-96-135 Page 5-2 reactor trip on turbine trip, the turbine bypass system, the steam dump system, or the Power Operated Relief Valves (PORVs).

Two cases are normally analyzed for this event: one challenging over-pressurization criteria, and one challenging the fuel design limit (MDNBR Case). 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, the DNBR for this event will continue to be bounded by the Loss of Coolant Flow event. Reanalysis of the event is not required for DNBR.

5.3 Anal sis Results The over-pressurization analysis of this event was performed at a Technical Spec'ification PCS flow rate of 345,000 gpm, and with 30'/o SGTP. The combination of decreased PCS flow and increased SGTP results in a decrease in the initial steam generator secondary pressure from 790 psia to 784 psia. The peak pressurizer pressure was calculated to be 2607 psia. Using this peak pressurizer pressure, the peak system pressure becomes 2714 psia at the outlet of the PCS pumps. This is below the criteria of 110/o of design pressure (2750 psia), which satisfies the overpressure design criteria.

The peak steam generator secondary pressure was calculated to be 1031 psia. This is below the criteria of 110~/o of design pressure (1100 psia), which satisfies the overpressure design criteria.

Table 5.1 on the following page gives a sequence of events summary for this 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.

EMF-96-1 35 Page 5-3 Table 5.1 Summary of Events for the Loss of External Load EVENT TIME (Sec)

Turbine Trip 0.00 Pressurizer Heaters on 0.00 Reactor Scram (Begin Rod Insertion) on High Pressurizer Pressure 3.59 Pressurizer Safety Valve Opens 4.01 Peak Power 4.06 Steam Line Safety Valves Open 5.35 Peak Primary Pressure 5 43 Peak Core Average Temperature 6.60 Peak Steam Dome Pressure 7,66

e EMF-96-1 35 Page 5-4 8000.0 2600.0 2000.0 1600.0 P4 1000.0 600.0 0.0 0.0 10A) iKO Time (sec)

Figure 5.1 Reactor Power Level for Loss of External Load

EMF-96-1 35 Page 5-5 O

26.0 20.0

>6.0 gl 10.0 g

8 6.0 0.0 0.0 6.0 10.0 16.0 20.0 26.0 SO.O Time (sec)

Figure 5.2 Core Average Heat Flux for Loss of External Load

1 EMF-96-1 35 Page 5-6 500.0 Gore Aeg Tealpmeture Core Inlet Temperature 590.0 Cold teg Temperature Hot Leg Temperature 680.0 570.0 E" 650.0

/ /

/ //

650.0

/ //

/

540.0 //

0.0 6.0 10.0 15.0 20A) SO.O Time (seo)

Figure 5.3 Reactor Coolant System Temperatures for Loss of External Load

e EMF-96-1 35 Page 5-7 2800.0 2800.0 d

2400.0 22008) 2000.0 1800.0 0.0 5.0 10A) 15.0 20Al 25.0 80.0 Time (sec)

Figure 5.4 Pressurizer Pressure for Loss of External Load

~

e'MF-96-1 Page 35 5-8 2.0

-0.0

-2.0 Total Baottvtty Doppler RacUvlty 0 Moderator Raottvtty I

l4 MO MO 0.0 6.0 iOA) 16.0 20.0 26.0 SO.O Time (sec)

Figure 5.5 Reactivities for Loss of Externai Load

EMF-96-1 35 Page 5-9 1100A) 1000.0 900.0 8008) 700.0 0.0 100 16.0 20.0 2KO SOA)

Time (sec)

Figure 5.6 Secondary Pressure for Loss of External Load

EMF-96-1 35 Page 6-1 6.0 REVIEW OF MECHANICALDESIGN The fuel mechanical design has been evaluated relative to the lower core flow/increased core coolant temperature and increased fast flux expected due to the increase in steam generator tube plugging from 25% to 30%. The results of this evaluation are as follows:

~ The increased tube plugging has no impact on the fuel assembly mechanical design.

Component and assembly strength is not affected by the changed operating conditions.

Corrosion remains bounded by fuel rod corrosion, which increases by only a few microns due to coolant temperature.

The most limiting fuel rod design criterion, transient stress, was recalculated with the lower flowfincreased core coolant temperature conditions. The calculation showed that cladding transient stress changes only slightly, meeting the design criterion. All other design parameters will remain essentially unchanged, and margins to design limits are not significantly degraded.

L EMF-96-1 35 Page 7-1

7.0 REFERENCES

1. Letter, R. J. Rodriquez (FPL) to T. M. Howe (SPC) "St. Lucie Unit 1 Cycle 14 Design Input For Analysis to Support 30% Steam Generator Tube Plugging," NF-96-206, May 23, 1996

2. St. Lucie Unit 1 C cle 14 Safet Anal sis Re ort FSAR Cha ter 15 Review and 99352, May 1996

3. Standard Review Plan for the Review of Safet Anal sis Re ort for Nuclear Power Plants, NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981.

4. Plant Transient Anal sis for St. Lucie Unit 1 Reactor, XN-NF-82-99, Exxon Nuclear Company, Richland, WA 99352, January 1983.

5. St. Lucie Unit 1 U dated Final Safet Anal sis Re ort, Florida Power and Light Company, Miami, FL 33174, July 1991.

6. St. Lucie Unit 1 Assessment of Radiolo ical and Rod Bow Effects for Increased Burnu, ANF-88-113(P), Advanced Nuclear Fuels Corporation, Richland, WA 99352, July 1988.

7. Steamline Break Anal sis for St. Lucie Unit 1, XN-NF-85-85(P), Exxon Nuclear Company, Richland, WA 99352, November 1985.

8. St. Lucie Unit Loss-of-Feedwater Transient with Reduced Steam Generator Low Level M-Richland, WA 99352, March 1994.

9. St. Lucie Unit 1 Cha ter 15 Event Review and Anal sis for 25'o Steam Generator Tube flu ~in, EMF-92-165, Siemens Power Corporation, Richland, WA 99352, March 1 983.

10. A Generic Anal 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.

11. St. Lucie Unit 1 Small Break LOCA Anal sis, EMF-92-148, Revision 1, Siemens Power Corporation, Richland, WA, 99352, May 1994.

e EMF-96-1 35 Page A-1 APPENDIX A The following NRC approved computer codes were used for new safety analysis supporting this report:

XN-NF-81-58(A), Revision 2, and Supplements 1 through 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nucl'ear Company, Revision 2 and Supplements 1 and 2 dated March 1984, Revision 2, Supplements 3 and 4 dated June 1990.

XN-NF-74-5(A) Supplements 1 through 6 and Revision 2, "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR),"

Exxon Nuclear Company, October 1986, XN-NF-75-21(A) Revision 2, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient Core Operation," Exxon Nuclear Company, January 1986.

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eY

EMF-96-135 Issue Date: g(3] /9Q St. Lucie Unit 1 Chapter 15 Event Review and Analysis for 30% Steam Generator Tube Plugging DISTRIBUTION R. A. Copeland, 21 K. M. Duggan, 35 R. C. Gottula, 36 T. M. Howe / FPL (5), 38 B. D. Stitt, 36 Document Control (2)

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