ML17223A556
| ML17223A556 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/09/1990 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17223A555 | List: |
| References | |
| NUDOCS 9003150489 | |
| Download: ML17223A556 (43) | |
Text
AT ACHME 1
PROPOSED TECHNICAL SPECIPICATIONS Marked-up Technical Specification Fages 2-4 3/4 3-15 3/4 3-17 9003i504S9 900309 PDR AOOCK 05000335 P
TABLE 2.2-1 REACTOR PROTECTIVE INSTRNENTATION TRIP SETPOINT LIMITS I
m FUNCTIONAL UNIT 1.
Power l.evel.-., High (1)
Four Reactor Coolant Pusps Operating 3,
Reactor Coolant Flow - low (1)
Four Reactor Coolant pyaps Operating TRIP SETPOINT Not Applicable
< 9,61$ above THERMAL POWER, with a winiest setpoint of l5X of RATIO THERNL R, and a maxim'f <'107.l5 of RATED THEeeu.
POWER.
> 95% of design reactor coolant flow with 4 pumps operating~
ALLOWABLE VALUES Not Applicable
< 9.61$ above THERMAL POWER, a
d a minimus setpoint of 15K of E
THERMAL POWER and a maxim'
< 107.0$ of RATED THERMAL POWER.
> 95% of design reactor coolant Slow with 4 pumps operating*
4.
Pressurizer Pressure High
< 2400 psia 5.
Contaireent Pressure - High
< 3.3 psig 6.
Steaa Generator Pressure - Low (2)
> 600 psia 7.
Steam Generator Water Level -Low WMC ihter level - each Q.
generator 8.
Local Power Density - High (3)
Trip setpoint adjusted to not exceed the limit lines of Figures 2.8-1 and 2.2-2 V
- Design reactor coolant flow with 4 pumps operating is 370f000 gpm.
< 2400 psia
< 3.3 psig
> 600 psia
@gal g/
(
~IV-.8% Mater Level - each aa generator Trip set point adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2.
TABLE 3. 3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUHENTATION TRIP VALUES" FUNCTIONAL UNIT 6.
LOSS OF POWER TRIP VALUE ALLOWABLE VALUES a.
(1) 4.16 kv bnergency Bus Undervoltage (Loss of Voltage)*
b.
4.16%v Emergency. Bus Undervol tage (Degraded Voltage) 2900
+ 29 volts with a 2900
+ 29 volts with a 1
+.5 second time delay 1
+.5 second time delay (1)
Undervol tage Device f1*
(2)
Undervol tage Device S2*
c.
480 volts Emergency Bus Undervol tage (Degraded Voltage)*
7.
AUXILIARYFEEDWATER (AFAS) a.
Hanual (Trip Buttons) b.
Automatic Actuation Logic c.
SG 1A 5 1B Level Low 8.
AUXILIARYFEEOMATER ISOLATION a.
Steam Generator dP-High b.
Feedwater Header High hP 3675'+ 36 volts with a 7 +
1 minute time delay 3592 + 36 volts with a 18 + 2 second time delay 429 + 5-0 volts with a 7 +
1 second time delay Not Applicable Not Applicable
~h)R D'ro
<275 psid
<150.0 psid 3675
+ 36 volts with a 7
+
1 minute time delay 3592
+ 36 volts with a 18 +
2 second time delay 429
+
5 -0 volts with a 7 +
1 second time delay Not Applicable Not AppliCab]e
~ 18.0'lo
-pe@ ~ RX428l PslD
- This specification will be effective prior to Cycle 7 restart.
9).QQ 5t.5PslD
~ L
~ ~I
~
I. ~
aw I
'a r-II t
I
~
llr~
I f'PAL.
II ~
)1 ~I ~
TABLE 3.3-5 Cont fnucd ENGINEERED SAFETY FEATURES
RESPONSE
INITIATINGSIGNAL AND FUNCTION 3.
Contafnment Pressure-Hf h
a.
Safety Igcctfon (ECCS) b.
Contafneent Isolatfon~
c.
Shfeld Bufldfng Ventflatfon Systea d.
Contafreent Fan Coolers e.
Feedwater Isolatfon 4.
Contafreent Pressure Hf h-Hf h a.
Contafment Spray 5.
Contafreent Radfatfon-Hf h a.
Contafreent Isolatfon~
b.
Shfeld Bufldfng Ventflatfon System 6.
S ae nerator P
s a.
Nafn Steam Isolatfon b.
Feedwater Isolatfon 7.
Refuel fn Mater Stora Tank-Los a.
Contafreant Suap Recfrculatfon 8.
Stcam Generator Level-Let a.
Auxf1fary Feedwater TINES RESPONSE TI S IN SECONDS
< 30.0 /19.5~
< 30.5~/20.5~
< 30.%/14.~
< 30.0 /17.~
< 60.0
< 30.0 /18.5~
< 30.5~/20.5~
< 30.0 /14.%
< 6.9
< 60.0
< 91.5
~ 205~
TABLE NOTATION Dfesel generator startfng and sequence loadfng delays fncluded.
~Dfesel generator startfng and sequence loadfng delays
~no
@eluded.
Offsfte power avaflable.
~Not applfcable to contafreent fsolatfon valve IM-18-1.
ST. LUCIE - UNIT 1" 3/4 3-17 Amendmant No. J7,$ T,H. 72
'ATTACHMENT 2 SAFETY ANALYSIS INTRODUCTION BACKGROUND This is a
request to revise Technical Specifications 2.2-1, "Reactor Trip Setpoints" and 3/4.3.2, Engineered Safety Features Actuation System Instrumentation Trip values for St. Lucie Unit 1.
On September 24, 1987, the Commission issued Amendment f23 to the St. Lucie Unit 2 Technical Specifications, which amended the Steam Generator Level-Low Reactor Protective System (RPS) and Auxiliary Feedwater Actuation System (AFAS) trip and allowable setpoints.
The intent of this license amendment proposal is to incorporate similar values in the St. Lucie Unit 1 Technical Specifications as presented in Attachment 1.
Operation of St.
Lucie Unit 1 with reduced Steam Generator (S/G)
Level-Low RPS and AFAS trip setpoints is desired for the following two reasons.
- First, by lowering the S/G level setpoint, larger fluctuations can be accommodated without causing an automatic reactor trip or unnecessary actuation of the Auxiliary Feedwater System (AFWS).
Since the RPS and AFAS setpoints are being reduced, the upper limit for AFWS response time is also being reduced to ensure adequate S/G inventory is available during design basis transients.
The second reason for the change is related to human factor concerns in that the proposed change will provide setpoints which are consistent with those previously approved for St. Lucie Unit 2.
The reason for changing the St. Lucie Unit 1 Auxiliary Feedwater Isolation allowable values is to make them consistent with those previously approved for St. Lucie Unit 2 in Amendment N28, dated March 22,
- 1988, and to provide additional assurance that the AFAS isolation logic properly identifies the faulted steam generator.
DESCRIPTION OF TECHNICAL SPECIFICATION CHANGES The proposed license amendment will revise Item 7 of Table 2.2-1, Reactor Protective Instrumentation Trip Setpoint
- Limits, of Specification 2.2.1, "Reactor Trip Setpoints,"
to reflect a
reduction in the Steam Generator Water Level-Low trip setpoint and allowable values.
The existing specification requires that the trip setpoint on S/G water Level-Low be
> 37.0% Narrow Range (NR) with an allowable value of > 37.04 NR.
The proposed change decreases the trip setpoint to > 20.54 NR with an allowable value of > 19.5>
NR.
l The proposed license amendment, will also revise Item 7.c of Table 3.3-4, "Engineered Safety Feature Actuation System Instrumentation Trip Values,
" of Technical Specification 3.3.2.1, "Engineered Safety Feature Actuation System Instrumentation,"
to reflect, a
reduction in'the trip and allowable value for the AFAS on S/G Water Level-Low. Item 7.c presently requires that the AFAS setpoint on Steam Generator Level-Low be set at
> 29.04 NR with an allowable value of > 28.54 NR. The proposed change would revise the AFAS trip setpoint on S/G Level-Low to
> 19.04 NR, with an allowable value of > 18.0%
NR.
A change is also being made to Item 8.a of Table 3.3-5, "Engineered Safety Features
Response
Times," to reflect the new AFWS response time for the S/G Level-Low signal.
Item 8.a presently specifies that the AFWS response time on S/G Level-Low is
> 205 and
< 600 seconds.
The proposed change reduces the AFWS response time upper limit on S/G Level-Low from 600 to 305 seconds.
The AFWS response time lower limit of > 205 seconds remains unchanged.
The proposed license amendment will also revise Item 8.a of Table 3.3-4, which presently specifies a S/G differential pressure-high trip value of < 275 psid with an allowable value of < 281 psid, to add a
lower limit of 89.2 psid to the allowable
- value, thus yielding an acceptable allowable range of 89.2 psid to 281 psid.
Additionally, the proposed license amendment will revise Item 8.b of Table 3.3-4, which presently specifies a
feedwater header differential pressure-high trip setpoint of < 150.0 psid with an allowable value of < 157.5 psid, to add a lower limit of 56.0 psid to the allowable value, thus yielding an acceptable value range of 56.0 psid to 157.5 psid.
DISCUSSION The RPS provides protection through the S/G Level-Low reactor trip to ensure against operation with insufficient secondary liquid inventory.
The AFAS is designed to automatically respond to decreasing steam generator coolant inventory.
The AFAS setpoint assures that the AFWS will be actuated when the actuation signal is reached.
The specified range for AFWS response time permits the operators to assess the post trip plant condition, without the concern of the automatic actuation of the AFWS within seconds of the trip.
The function of the Auxiliary Feedwater Isolation trip and allowable values for steam generator and feedwater header high differential pressure is to isolate auxiliary feedwater flow to a faulted S/G under conditions of either a
Steam Line Break or a
Feedwater Line Break.
These two functions were implemented on St.
Lucie Unit 1 in response to NUREG-0737.
The previously described proposed changes were evaluated to determine the impact on the existing Safety Analysis, particularly,
I to assess the impact on the licensing bases defined in Final Safety Analysis Report (FSAR) Chapter 15 and the impact on the AFWS design bases defined in Chapter 10.
Im act of the Pro osed
.Chan es on Final Safet Anal sis Re ort Cha ter 15 Accident Anal ses A review of the FSAR Chapter 15 events for St.
Lucie Unit 1 was performed to assess the effects of reducing the S/G Level-Low RPS, AFAS trip setpoints and reducing the upper limit of the AFWS response time, in order to determine which events needed to be reanalyzed.
A summary of the category of events considered and their disposition is presented in Table 2-1.
The review shows that the only category of events that is impacted is the "Decrease in Heat Removal by the Secondary System."
In this
- category, event initiators cause the loss of the secondary heat sink and a
subsequent primary system heatup.
Several reactor trip functions are available to mitigate the consequences of these events.
In terms of challenging the primary vessel pressure limits the most limiting event is in the Loss of External Load (LOEL) event.
All the events in the "Decrease in Heat Removal by the Secondary System" category which result in rapid RCS pressurization are bounded by the LOEL event.
The High Pressurizer Pressure trip is the primary controlling trip for this event.
In terms of challenging the adequacy of S/G liquid inventory, the most limiting event in the "Decrease in Heat Removal by the Secondary System" category is the Loss of Normal Feedwater Flow (LONF). This event is initiated by a malfunction of the feedwater system, resulting in a gradual reduction of S/G liquid inventory.
The inventory reduction eventually causes a decrease in the heat removal rate from the primary system by effectively decreasing the heat transport area as the secondary liquid level drops.
The S/G Water Level-Low trip is the primary controlling trip for this event.
The reduction of the S/G Water Level-Low trip RPS and AFAS setpoints and the AFWS response time impact the results of the LONF analysis.
Therefore, reanalysis of this event with respect to minimum S/G inventory requirements was required.
The reanalysis is presented in Attachment 4.
The results demonstrate that the proposed setpoint changes are adequate to maintain a secondary heat sink throughout the LONF event, as described in FSAR Section 15.2.8.
A sensitivity calculation, in which no auxiliary feedwater was delivered to the steam generator, indicated that dryout would occur at 700 seconds.
The proposed AFWS response time of 305 seconds is the same as the St. Lucie Unit 2 and is well within the time frame that the Operators have, i.e.,
700 seconds, to verify automatic AFW pump start or manually initiate auxiliary feedwater.
Therefore, it can be concluded that the proposed RPS S/G Level-Low trip setpoint ensures that sufficient water inventory exists in the S/G's at the time of the trip to provide a margin of more than 10 minutes before AFW is required as stated in the Bases of the Technical Specifications.
The Auxiliary Feedwater Isolation lower bound allowable values for S/G and feedwater header differential pressure high will ensure that negative instrument errors will not cause the isolation logic trip value to encroach upon the lower end of the instrument range.
A review of the instrument loops was performed and it was confirmed that the St. Lucie Unit 2 allowable values are applicable to St.
Lucie Unit 1.
A review of the appropriate accident analyses was also performed to ensure that the proposed changes to the auxiliary feedwater isolation trip and allowable values for S/G and feedwater header high differential pressure would not impact the analysis results.
No accident analyses were identified that take credit for the auxiliary feedwater isolation feature.
The Steam Line Break accident presented in Section 15.4.6 of the St. Lucie Unit 1 FSAR conservatively assumes that auxiliary feedwater is admitted into the affected steam generator and is manually isolated ten minutes into the event.
Feedwater system pipe breaks were considered a
cooldown event in the licensing basis for St. Lucie Unit 1 (FSAR Section 15.2.8.3.2) and as such are bounded by the Steam Line Break event analysis.
Im act of Pro osed Technical S ecification Chan es on FSAR Cha ter 10 Anal ses The design bases of the AFWS and the evaluation of the system are presented in FSAR Section 10.5.
The AFW system evaluation was performed based on the Standard Review Plan (SRP)
Section 10.4.9 and the Branch Technical Position (BTP) 10-1 guidelines.
In the original response to the NUREG-0737 requirements, Florida Power
& Light concluded that, of the ten transients which were initially evaluated, the two most limiting were a Loss of Main Feedwater (LMFW) concurrent with an AFWS High Energy Line Break (HELB) and a Loss of Offsite AC Power concurrent with an AFWS HELB.
These transients were re-evaluated to access the impact of the reduced S/G Water Level-Low for RPS and AFAS trip and allowable setpoints as well as the reduced AFWS response time.
A summary of the evaluation of these plant transients and their disposition is presented in Table 2-2.
Based on the review of these transients, the following two events were analyzed:
(1)
LMFW with AC Power available (2)
LMFW with Loss of Offsite AC Power.
Both of these analyses assumed a concurrent AFWS HELB.
The LMFW with AC power available is the most limiting case with regard to minimum S/G inventory since the reactor coolant pumps (RCPs) are available to add additional heat to the RCS, whereas in the LMFW with Loss of Offsite AC Power the RCPs do not continue to add heat to the primary coolant.
The results for the limiting case of LMFW with AC power available concurrent with AFWS HELB at the steam-driven pump discharge, demonstrate that the AFWS at St. Lucie Unit 1 is adequate to maintain a secondary heat sink for up to 30 minutes after the S/G Level-Low trip.
After 30 minutes it would be required that one RCP in each coolant loop be tripped by the Operators in order to reduce the addition of the RCP heat input into the
- system, and thus to recover steam generator levels as indicated in the Emergency Operating Procedures for St. Lucie Unit 1.
CONCLUSIONS Based on the results of the event reviews and the results of the reanalysis for both FSAR Chapters 10 and 15, it can be concluded that the incorporation of the proposed changes can be implemented since the important safety parameters remain within the appropriate acceptance criteria and there is no reduction in the margin of safety.
5 K
TABLE 2-1 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON THE FSAR CHAPTER 15 EVENTS EVENT CATEGORY IMPACT DISPOSITION A.
Increase in Heat Removal by the Secondary System.
(SRP 15.1)
The events in this category result in RCS cooldown due to increase in steam demand.
The proposed changes have no impact on either the event initiators or the controlling reactor trips.
Prior analyses of record bound the proposed changes.
No reanalysis is necessary.'.1 Steam System Piping Failures Inside and Outside of Containment (SRP 15.1')
Secondary blowdown and depressurization cause RCS cooldown that could result in possible return to power.
The SG Low-Level trip RPS and AFAS setpoints are not credited in this analysis.
This analysis assumes that auxiliary feedwater is delivered 180 seconds into the transient.
This assumption defines the lower AFAS response time Technical Specification value of 205 seconds (180 second analysis assumption plus 25 second AFAS timer error).
Since there is no change in the lower AFAS response time value, there is no impact on this analysis.
Changes are bounded by previous analysis of records.
No re-analysis is necessary.
TABLE 2-1 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON THE FSAR CHAPTER 15 EVENTS EVENT CATEGORY IMPACT DISPOSITION B.
Decrease in Heat Removal by the Secondary System (SRP 15.2)
The proposed changes impact the Loss of Normal Feedwater Flow event.
This event is analyzed primarily to evaluate minimum SG liquid inventory.
The controlling reactor trip in this event is the SG Low-Level, thus the change of the RPS trip setpoint impacts the results of the analysis.
Reanalysis of the Loss-of Normal Feedwater Flow event (SRP 15.2.7, FSAR 15.2.8) was necessary.
The remaining events in this category are mitigated by different reactor trips and therefore are bounded by existing analyses of record.
B.l Feedwater System Pipe Breaks Inside and Outside Containment (SRP 15.2.8)
This event is a cooldown event in the licensing basis for the plant.
The Feedwater Pipe Break event is bounded by the Steam Line Break event since the area for flow in a broken feedwater pipe is less than that of a
severed steamline resulting in a more benign event. "
The proposed changes have no impact on this event.
This event is bounded by SRP 15.1.5 event (FSAR 15.4.6),SLB analysis.
TABLE 2-1 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON THE FSAR CHAPTER 15 EVENTS EVENT CATEGORY IMPACT DISPOSITION C.
Decrease in Reactor Coolant System Flow Rate (SRP 15.3)
The events in this category are initiated by Reactor Coolant Pump malfunction and result in a
reduction of RCS flow.
These transients are mitigated by VHPT, Hi Pressure Low RCS Flow trips.
The proposed changes have no impact on the previous analyses.
Prior analyses of records bound the proposed changes.
No reanalysis is necessary.
D.
Reactivity and Power Distribution Anomalies (SRP 15.4)
E.
Increase in Reactor Coolant Inventory (SRP 15.5)
The events in this category result in a
power increase and are controlled by
- VHPT, LPD
- LSSS, or TM/LP trips.
The proposed changes have no impact on the previous analyses.
Bounded by previous analysis results.
These events are not part of the licensing basis for St. Lucie Unit 1.
TABLE 2-1 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON THE FSAR CHAPTER 15 EVENTS EVENT CATEGORY IMPACT DISPOSITION F.
Decrease in Reactor Coolant Inventory (SRP 15.6)
Prior analyses do not credit SG Low-Level
Prior analyses of record bound the proposed changes.
G.
Radioactive Releases from a system or component (SRP 15.7)
None of these events occur as a
direct consequence of operation of the reactor.
SG Low-Level trip RPS and AFAS setpoints as well as AFAS response time are not
- credited, thus change of these setpoints do not impact previous conclusions.
Prior conclusions bound the proposed changes.
H.
Asymmetric Events The events in this category are controlled by the Asymmetric Steam Generator Transient Protective Trip Function (ASGPTF) be fore asymmetries can be developed, thus the symmetric events remain bounding.
Existing'nalyses bound the proposed changes.
TABLE 2-2 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES'N FSAR CHAPTER 10 ANALYSES Transient Description of Transient/
Accident Condition Evaluation-Disposition 1 ~
2 ~
3 ~
Loss of Main Feedwater (LMFW) concurrent with a High Energy Line Break (HELB).
Loss of Main Feed with Loss of Offsite AC Power concurrent with a HELB.
LMFW with Loss of Onsite and Offsite AC Power.
This transient was reanalyzed to assess impact of reducing the SG Low Level trip setpoint and AFAS setpoint.
The results demonstrate that the AFWS at St.
Lucie Unit 1 is adequate to maintain a secondary heat sink up to 30 minutes after the SG Low Level trip.
After 30 minutes it would be required that one Reactor Coolant Pump in each coolant loop be tripped by the Operators to recover steam generator levels.
This transient was reanalyzed and results in a condition that is less limitingthan transient gl because there are no Reactor Coolant Pumps operating to supply additional heat input.
Bounded by transient
$ 1.
This transient was not assumed to occur concurrent with HELB in the AFWS. Since loss of onsite diesel generators were the assumed failures, this resulted in the availability of the steam-driven AFW pump with sufficient capacity to remove decay heat.
Bounded by transient g1.
TABLE 2-2 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON FSAR CHAPTER 10 ANALYSES Transient Description of Transient/
Accident Condition Evaluation-Disposition 4
5 ~
6.
7 0 8.
Turbine Trip with and without bypass concurrent with the AFWS HELB.
Main Steam Isolation Valve Closure concurrent with the AFWS HELB.
Main Feed Line Break Main Steam Line Break Small Break LOCA This transient is a
less limiting transient than g1 since offsite AC power is available to operate the MFW pumps.
This transient is thus bounded by the results of transient gl.
This transient is a
less limiting
'transient than 41 since offsite AC power is available to operate the MFW pumps.
This transient is thus bounded by the results of transient g1.
This transient is not assumed concurrent with an AFWS HELB and therefore the entire capacity of the AFWS is available.
This transient is not assumed concurrent with an AFWS HELB and therefore the entire capacity of the AFWS is available.
This transient is not assumed concurrent with an AFW HELB and therefore the entire capacity of the AFWS is available.
TABLE 2-2 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON FSAR CHAPTER 10 ANALYSES Transient Description of Transient/
Accident Condition Evaluation-Disposition 9
Loss of Offsite AC Power with AFW HELB In this transient the limitingauxiliary feedwater line break may occur at two different locations.
The first potential break was modeled at the discharge of the turbine-driven pump.
This case is bounded by transient 42 which is less limiting than transient gl.
The second potential auxiliary feedline break location was modeled at the discharge of the second motor-driven AFW pump "B" and assuming a failure of "A"battery, (failure of"A"battery willresult in failure of "A"-ACmotor driven pump and the AB-DC tie bus to A bus).
In this case, the operators have 11.7 minutes (SG dryout time established in Section 15.2.8) to initiate flow via the turbine-driven AFW pump by transferring electrical DC loads from the dead "A" bus to the energized "B" bus.
This transfer as presented in Section 10.5 of the FSAR is conservatively assumed to take 5
minutes.
Since the capacity of a steam-driven pump is greater than either of the two motor driven AFW pumps, SG inventories begin to recover immediately.
Thus this case is less limiting than transient g1.
TABLE 2-2 IMPACT OF PROPOSED TECHNICAL SPECIFICATION CHANGES ON FSAR CHAPTER 10 ANALYSES Transient Description of Transient/
Accident Condition Evaluation-Disposition 10.
Plant cooldown concurrent with an AFWS HELB.
This is a less limitingtransient than
$1.
Since offsite power is available to operate the MFW pumps, this transient is bounded by transient gl.
E
,ATTACHMENT 3, DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION t
The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission s regulation 10 CFR 50.92, which states
.that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously evaluated; or (3) involve a
significant reduction in a margin of safety.
Each standard is discussed as follows:
(1)
Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to lower the Steam Generator (S/G) Level-Low trip Reactor Protective System (RPS) and Auxiliary Feedwater Actuation System (AFAS) setpoints can potentially reduce the likelihood of an unplanned reactor trip or AFAS initiation by allowing larger fluctuations in the S/G water level.
Additionally, the proposed change to the AFWS response time reduces the maximum response time and thus reduces challenges to steam generator (S/G) integrity under the condition of a plant trip.
The only event relying on the reactor trip on low S/G level and the AFWS response time for mitigation is the Loss of Normal Feedwater Flow from the standpoint of minimum S/G inventory requirements; Section 15.2.8 in the Final Safety Analysis Report (FSAR).
The results show that the operators have 11.7 minutes to verify auto start or, if necessary, to manually initiate auxiliary feedwater flow before S/G dryout occurs.
Therefore, the proposed RPS S/G Level-Low setpoint ensures that sufficient water inventory exists in the S/G's at the time of the trip to provide a margin of more than 10 minutes before AFW is required as stated in the Bases of the Technical Specifications.
The AFWS response time of 305 seconds is well within the time-frame demonstrated as acceptable in the Bases of the Technical Specifications.
The proposed S/G Level-Low RPS trip and AFAS setpoints have been established such that, they ensure actuation of those functions and include conservative allowances for instrumentation uncertainties.
The reanalysis of the Loss of Normal Feedwater Flow event is performed accounting for the
most adverse combination of uncertainties of the proposed setpoints ensuring that the required instrumentation remains on scale for system actuation.
The proposed changes to the auxiliary feedwater isolation trip setpoint and the allowable values for S/G and feedwater header high differential pressure provide additional assurances that the AFAS Auxiliary Feedwater Isolation logic would properly identify a faulted steam generator under certain accident scenarios.
The proposed change is a conservative change to the Technical Specifications in that it provides a tighter band for the Auxiliary Feedwater Isolation allowable trip values.
The proposed change does not impact accidents previously evaluated for St. Lucie Unit 1.
The Auxiliary Feedwater System'as evaluated against 'the Standard Review
- Plan, Section 10.4.9 and Branch Technical Position 10-1 guidelines.
From this evaluation, it was concluded that the Auxiliary Feedwater System continues to meet the acceptance criteria of the Standard Review Plan and Branch Technical Position.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Operation of the facility in accordance with the proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously evaluated.
No new accident indicators are created by the changes needed to support.
reduction of the S/G Level-Low setpoints for actuation of a reactor trip and initiation of the
- AFAS, as well as the changes needed to support a reduction of the AFWS response time.
Additionally, no new accident initiators are created by the changes needed to support a revision to the Auxiliary Feedwater Isolation,trip setpoint and allowable values for steam generator and feedwater header high differential pressure.
The events reanalyzed provide assurance that operation of AFAS with reduced RPS trip and AFAS setpoints, as well as a
different AFWS response time, produces acceptable results.
Finally, the changes do not result in any physical change to the plant or method of operating the plant from that allowed by the Technical Specifications.
Therefore, the proposed Technical Specification changes do not create the possibility of a
new or different kind of accident.
(3)
Operation of the facility in accordance with the proposed amendment would not involve significant reduction in a margin of safety.
The reduced S/G Level-Low RPS and AFAS trip setpoints and the AFWS response time change have been evaluated for their impact on the current safety analysis.
The results of the new analyses for these transients impacted by the proposed changes remain within the appropriate acceptance criteria.
The changes made to the isolation trip setpoint and the differential pressure values do not have any effect on the results of any previously analyzed events.
Therefore, the existing margin of safety is preserved.
Based on the above, we have determined that the proposed amendment does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously evaluated; or (3) involve a
significant reduction in a margin of safety.
ATTACHMENT 4 ST ~
LUCIE UNIT 1 LOSS-OF-FEEDWATER TRANSIENT WITH REDUCED STEAM GENERATOR LOW-LEVEL TRIP SETPOINT
4.0 SA HA T NA YS This section describes calculations performed to fulfill requirements given in Section 15.2.7 of the SRP for a Loss-of-Feedwater event.
Chapter 15 events are analyzed using conservatively biased plant data for initial conditions.
Section 15.2.7 of the SRP requires a limiting single failure, but does not require it to be concurrent with.a limiting auxiliary feedwater line break, as in the Chapter 10 analysis.
In this analysis, two cases were analyzed.
In the base
- case, the single failure was assumed to be the turbine driven AFS pump.
This is conservative because the turbine driven pump has the greatest feedwater capacity.
This assumption leaves both motor driven AFS pumps operable with one motor driven AFS pump feeding each steam generator.
Because the most conservative results are obtained with the RCP running, only the offsite power available case was considered in this analysis.
A sensitivity calculation was also performed which was identical to the base calculation except that it was assumed the automatic initiation of auxiliary feedwater failed.
The purpose of this calculation was to determine how much time the operators have'efore steam generator dryout occurs to initiate flow from an AFS pump.
This calculation supports a Technical Specification limit
~
of 10 minutes for operator action to occur.
41 S0 -F W
0 S
This event was initiated by a total loss of all main feedwater.
Figures
- 4. 1 through 4.7 show various parameters from both the primary and secondary system
- response, with 0.0 seconds being the time of the SG low level trip.
Table
- 4. 1 lists the initial conditions for the transient, which have been biased in a conservative direction.
Table 4.2 presents the major assumptions for the analysis.
Table 4.3 presents the sequence of events for the transient.
The reactor tripped on low steam generator level 34 seconds after.transient initiation.
Following the reactor trip,.the turbine
- tripped, closing the steam flow control valves.
The Main Steam Safety Valves opened briefly to
P J
relieve the rapid secondary pressure increase.
The steam bypass control system then automatically activated and controlled the steam generator pressures to 910 psia.
This steam flow controlled both primary and secondary system temperatures as the steam generator inventory decreased.
The AFS began injecting water from both motor driven pumps to the steam generators at 330 seconds after the SG low level trip.
It was assumed that the turbine driven AFS pump was unavailable due to a single failure.
The flow from both motor driven AFS pumps was more than adequate to remove decay heat. Therefore, SG liquid inventories imaediately began to increase, as shown in Figure 3. 1.
The plant response was well behaved throughout the transient.
The minimum SG liquid inventory was 47,950
- ibm, and occurred at 330 seconds.
At the end of the calculation, 1000 seconds, the total liquid inventory in the steam generators was 68,410 ibm, and increasing.
The results of this analysis show that the Auxiliary Feedwater System is adequate to remove decay heat from the primary system and assure that the specified acceptable fuel design limits are not exceeded.
A sensitivity calculation was performed which was identical to the calculation described above,. except that no auxiliary feedwater was injected into the steam generators.
This calculation was performed to determine how much time the operators have to initiate flow from an AFS pump before steam generator dryout occurs.
Figures 4.8 and 4.9 show the steam generator liquid inventories and SG primary outlet temperatures respectively.
At 700 seconds, the SG outlet temperatures began to rise
- rapidly, indicating that heat transfer from the primary to the secondary systems was essentially lost.
This is a good indication of the effective steam generator dryout time.
Tabl 4.
Plant nitial Conditio For FSAR Cha ter 15 Anal sis PA HTR Core Power Core Inlet Temperature Pressurizer Pressure Pressurizer Liquid Level Primary System Loop Flow Rate Moderator Temperature Coefficient Doppler Coefficient Doppler Multiplier Steam Generator Pressure Steam Generator Liquid Level Steam Generator Secondary Mass Inventory Auxiliary Feedwater Temperature (Corresponds to enthalpy of 80.
BTU/ibm)
~VA U 2754 MW.t 551 'F 2297 psia
- 65. 5'1e 417,364 gpm 2.4 PCM/'F
-0.8 PCM/'F 1.0 904 psia 65% Narrow Range 141,056 ibm 110 'F
4.
A i
Assum tions For FSAR Cha ter 5 A al sis t
1 f
11 1
f 5
2.
t 5
tt ly t t ATAA Tip 5tp it, 15.55 1
Nar row Range SG level.
Actual value of RPS
- trip, with uncertainties, is 16.5%.
3.
S ri Set oi 15.3% of Narrow Range SG Level.
This value included all necessary uncertainties.
4.
w G
eve 15.25 ft.
The reference elevation for 0.% narrow range level is 25.09 ft above the top of the tube sheet.
5.
e re Cont 1
- PORVs, pressurizer
- heaters, and pressurizer spray systems were assumed to be
- operable, and to function at nominal setpoints.
Primary safety valves were assumed to open at the nominal pressure
- setpoint, 2500 psia.
6.
s e
1:
Steam dump and bypass systems were assumed to be operable in automatic
- mode, and opened at nominal setpoint pressures; Atmospheric dump valves are manually operated, and were assumed to be not available.
Secondary safety valves were assumed to open at 990 psia.
7.
e ratio All four Reactor Coolant Pumps (RCP) were on throughout the transient.
8.
w w
a The flow rate from both motor driven auxiliary feedwater pumps was delivered to both steam generators.
A 9'f measured flow curve was used to set the flow rates.
The initiation of auxiliary feedwater was delayed 330 seconds after the reactor trip.
9.
P d w S t m
No primary system charging or letdown flows were assumed.
Tabl 4.
A 1
A um tio For F
A ha ter 5 Ana i
cont.
10.
nvironmental h t losses:
No energy was assumed to be dissipated in either the primary or secondary systems to metal masses in the
- system, or to the containment environment.
This is a conservative assumption because it maximizes the energy that has to be removed by steam flow out of the generators.
11..
Stea Generator Tube Plu in evel:
It was assumed that no tubes are plugged in the steam generators.
This is a conservative assumption because it maximizes the heat transfer from the primary to the secondary systems.
12.
Reactor Kinetics Feedback and eca Heat Calculation:
Reactor kinetics feedback due to changing primary system conditions prior to reactor scram was modeled using the PTSPWR2 code.
After reactor
- scram, the ANS 1979 Standard decay heat formulation was applied, including the contribution due to actinide decay.
A 2
sigma uncertainty was applied to the decay heat calculations.
3 4
43 fE f
TARC~t5A
~Time se 0.0 0.9 1.2 330.
1000.
vent Scram signal on low SG level Reactor tripped Steam flow valves began to close Auxiliary feedwater to SGs initiated.
Minimum total SG inventory reached, 47,950 ibm Transient calculation
- ended, total SG liquid inventory:
68,410 ibm
~8 X
~
X CO CI 8
LEGEND 1-SG1 2-SG2 I 8 CO lA lC X:
8 1
1 1
1 0.00 2.00 4.00 6.00 TINE (SECONDS) 8.00 10.00 12.00 (10~~ 2)
Figure 4.1 SG Secondary Liquid Inventory, FSAR Chapter 15 Analysis
Cl Cl Cl Ol Cl LEGEND 1-SG1 2-SG2 43> Cl fag
~ ~
CO CI CV 0.00 2.00 4.00 6.00 TII1E (SECONDS) 10.00 12.00 (10xx 2)
Figure 4.2 SG Downcomer Liquid Level, FSAR Chapter 15 Analysis
8 X
~
X III Cl LEGEND 1-LOOP 1
2-LOOP 2 hl l?
Q Pe III lC fQ
~a
~ III 1
2 1
2 1
2 1
2 1
2 1
CI 0.00 2.00 4.00 6.00 TINE (SECONDS) 8.00 10.M 12.00 (1OIIII 2)
Figure 4.3 SG Primary Inlet Temperature, FSAR Chapter 15 Analysis
LEGEND 1-LOOP 1
2-LOOP 2 8
~
~
~ th b3 Cl 43 le L) f-s CI lA 1
2 1
2 1
2 1
2 1.
2 2
0.00 4.00 6.00 TINE (SECONDS) 8.00 10.00 12.00
(]Oxx 2)
>g<<e 4.4 SG Primary Outlet Temperature, FEAR chapter 15 Analysis
X gp XO CO 8
C3 8 CL1 CA E g
-1 d lQ CD LEGEND 1-CORE FLON LA 0.00 2.00 4.00 6.00 TIME (SECONDS) 8.00 10.00 12.00 (10xx 2)
Figure 4.5 Core Mass Flow Rate, FSAR Chapter 15 Analysis
O Cl CO e4 Cl CO LECENO 1-PZR LEVEL f-l Cd Cd o C
c4 Cd> a Cd
~,
CD CCl CI 0.00 2.00 4.00 6.00 TINE (SECONDS) 8.00 10.00 12.00 (10lClC 2)
Figure 4.6 Pressurizer Liquid Level, FSAR Chapter 15 Analysis
8 x~
A 8
CC a-8
~ Ol LL CA g 0
8 LEGENO 1-PZR PRESS C7ill al 0.00 2.00 4.00 6.00 TIME (SECONDS) 8.00 10.00 12.00 (10vx 2)
Figure 4.7 Pressurizer
- Pressure, FSAR Chapter 15 Analysis
(
~ s
.T l~
8 X
~
X CO Cl LEGEND 1
S81 2-SG2
$ 8 CA CAKs 0.00 2.00 4.00 6.00 TIIiE (SECONOS) 2 1
6.00 10.00 12.00 (1Qvx 2)
Figure 4.8 Secondary Liquid Inventory, FSAR Chapter 15 Analysis With No Auxiliary Feedwater
1$
C9 hd-I
~ lO
~$
0K 1-LOOP 1
2-LOOP 2 Cl LA 1
2 1
2 1
0.00 2.00 4.00 6.00 TINE tSECON05) 8.00 10.00 12.00 (1O)cx 2)
Figure 4.9 SG Primary Outlet Temperature, FSAR Chapter 15 Analysis With No Auxiliary Feedwater
0 t,+
'I9
~,
4
' +
TL'
'h u.)
'