ML17222A559
| ML17222A559 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/20/1988 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17222A558 | List: |
| References | |
| IEB-88-047, IEB-88-47, NUDOCS 8810270059 | |
| Download: ML17222A559 (84) | |
Text
ATTACHMENT 1 ST.
LUCIE UNIT 2 CEA DROP TIME MARKED-UP TECHNICAL SPECIFICATION PAGE 3/4 1-24 8810270059 881020 PDR ADOCK 05000389 l
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REACTIIi/IrY CONTROL SYSTEMS CEA OROP TIME r LIITING CQNOITION F'"R OP RATION
- 3. 1.3.4 The individual full-length (shutdown and regulating)
CEA dro time from a fully withdrawn pos'ition, shall 5e less than or equal to ++ secon s
from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90K inser:ion position with:
a.
T greater than or equal to 515~F, and avg All reactor coolant pumps operating.
APPLICABILITY:
MOOES 1
and Z.
ACTION:
a.
With the drop time of any full-length CEA determined to exceed the above limit:
l.
If in MOOE 1 or 2, be in at least HOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or Z.
If in MOOE 3, 4, or 5, restore the CEA drop time to within the above limit prior to proceeding to MOOE' or 2.
b.
With the CEA drop times within limits but determined at less than full reacto~ coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coo1ant pump combination operating at the time of CEA drop time determination.
~ 3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:
a.
For all CEAs following each removal and installation of the reactor vessel
- head, b.
For specifically affected individuals CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific
- CEAs, and c.
At least once per 18 months:
ST.
LUCIE - UNIT 2 3/4 1-24 Amendment No. 8
ATTACHMENT 2 Safet Anal sis Introduction A change in the maximum allowable control element assembly (CEA) drop time for the St. Lucie Unit 2 from the current value of 2.7 seconds to a value of 3.1 seconds is proposed.
This change is proposed to Limiting Condition for Operation 3.1.3.4, which specifies the maximum allowable CEA drop time.
- Recently, at Arkansas Nuclear One Unit 2, Arkansas Power and Light observed (NRC Information Notice No. 88-47) a small, but significant, increase in the CEA drop time when the CEA's were de-energized together instead of separately.
The CEA drop times were observed to increase by approximately 0.25 seconds.
Because a similar increase in the measured CEA drop times at St. Lucie Unit 2 may result in exceeding the current Technical Specification limit, Florida Power and Light will demonstrate the acceptability of longer CEA drop times.
To bound the expected increase in measured CEA drop times, an evaluation was performed to determine the impact on all Design Basis Events (DBE's) for a 0.4 second increase in the CEA drop time.
This was determined by analyzing the impact on the most limiting events with respect to the fuel and plant safety limits.
The results of this re-analysis were then shown to bound the consequence of the remaining DBE's.
A summary of the re-analysis, and the results, is provided below.
Discussion The CEA drop time was assumed to increase by 0.4 seconds from the previously assumed 0.34 seconds to 0.74 seconds as a result of an increase in the assumed Control Element Drive Mechanism (CEDM) holding coil delay time.
Large inductance coils around the CEDM magnetically hold the CEA s in position.
When a scram signal is
- received, these holding coils are de-energized.
- However, because of the large currents passing through these coils, there is a time delay associated with the decay of the magnetic field.
After the CEDM holding coil decay delay time, the CEDM s physically disengage and the CEA's drop into the core.
The DBE's have been re-evaluated against the currently approved analyses of record as presented in the St.
Lucie Unit 2 Updated Final Safety Analysis Report (UFSAR).
These events can be grouped in the following categories:
2.
3.
4.
5.
Increase in heat removal by the secondary system Decrease in heat removal by the secondary system Decrease in reactor coolant flow rate Reactivity 6 power distribution anomalies Decrease in reactor coolant system inventory 6.
Loss of coolant events An increase in CEA rod drop time by 0.4 seconds has an impact primarily on those events which (a) involve a rapid approach to a
C i4 I'
safety limit during the same time-frame as the scram and/or (b) the event involves a rapid approach to a Specified Acceptable Fuel Design Limit (SAFDL)
(minimum DNBR) during the first part of the scram insertion.
With respect to the various acceptance criteria for DBE s, certain events are more limiting than others.
These events were identified and re-analyzed.
The remaining events were evaluated using, as a
- basis, the impact of the increased holding coil delay time on the limiting events.
The results and conclusions of the re-analysis are presented below.
1.0 Increase in Heat Removal b
the Secondar S stem 1.1 Inside Containment Steam Line Break Pre-Tri Power Excursion The Pre-Trip Steam Line Break (SLB) event (with a Loss of A.C.
Power) is the limiting event with respect to reaching the minimum absolute DNBR value and the greatest amount of predicted fuel failure fraction.
To support an increase in the CEA drop time, a re-analysis was performed.
Two analysis assumptions were changed when compared to the analysis of record presented in the St.
Lucie Unit.
2 UFSAR:
1) increased CEDM holding coil delay time from 0.34 seconds to 0.74 seconds 2) time at which Loss of AC Power (LOAC) occurs.
The Analysis
1
'4
of Record assumed an overly conservative time for LOAC, occurring concurrently with the RPS trip on high containment pressure.
Sensitivity studies have shown that relaxing the time for the RPS trip high containment pressure by 3
seconds is a
more realistic (but still conservative) assumption.
Attachment 4 provides a thorough discussion of the re-
- analysis, the assumptions and results, following the UFSAR format.
The results of the re-analysis show a
predicted minimum DNBR of 0.782.
The analysis of record predicted a minimum DNBR of 0.783.
In both cases, the predicted fuel failure fraction is less than 10%. It can be concluded, based on the results of the re-analysis, that a
eoolable geometry is maintained since the predicted fuel failure fraction is less than 10<.
An additional calculation was performed to evaluate the impact on the minimum DNBR due only to the increased CEDM holding coil delay time.
The results of this analysis show a
34 degradation in minimum DNBR due to this analysis assumption change.
This evaluation will serve as reference to determine the impact on the minimum DNBR due to the increased CEDM holding coil delay time for less limiting events.
Site boundary doses for this event are bounded by the doses obtained in the outside
I I
V 'I'
containment SLB
- which, as discussed
- below, is not impacted by the proposed amendment.
Outside Containment Steam Line Break Post-Tri Power Excursion from Full Power The 0.4 second increase in the CEDM holding coil delay has a negligible impact on the minimum DNBR attained during this event since the time at which the CEA breakers are predicted to open is 4.3 seconds and the time of minimum DNBR is 119.6 seconds; at which time the DNBR value is well above the 1.3 SAFDL.
Post-Tri Steam Line Break from Zero Power Increasing the CEDM holding coil delay by 0.4 seconds has a negligible impact on the minimum DNBR attained during this event since the time of trip is 3.5 seconds and the time of minimum DNBR is 209.9 seconds; at which time the DNBR value is well above the 1.3 SAFDL.
Decrease in Feedwater Tem erature The increase in holding coil delay is non-conservative as it delays a trip and results in a lower DNBR.
The re-analysis of the Pre-Trip Steam Line Break event
indicates that the decrease in minimum DNBR is approximately 34 with respect to a 0.74 second versus a
0.34 second holding coil delay.
Applying this DNBR degradation to the Decrease in Feedwater Temperature event is conservative since the SLB event is a much more severe increase in heat removal event.
The last analysis of this event documents a
minimum DNBR of 1.34.
Accordingly, the expected decrease in minimum DNBR would result in a value well above the SAFDL of 1.28.
Increase in Feedwater Flow The increase in holding coil delay is non-conservative as it delays the trip and results in a lower DNBR.
The re-analysis of the Pre-Trip Steam Line Break event indicates that the decrease in minimum DNBR is approximately 3'~ with respect to a 0.74 second versus a
0.34 second holding coil delay.
Applying this DNBR degradation to the Increase in Feedwater Flow event is conservative since the SLB event is a much more severe increase in heat removal event.
The last analysis of this event documents a
minimum DNBR of 1.63.
Accordingly, the expected decrease in minimum DNBR would result in a value well above the SAFDL of 1.28.
Increase in Main Steam Flow The increase in holding coil delay is non-conservative as it delays the trip and results in a lower DNBR.
The re-analysis of the Pre-Trip Steam Line Break event indicates that the decrease in minimum DNBR is approximately 3% with respect to a 0.74 second versus a
0.34 second holding coil delay.
Applying this DNBR degradation to the Increase in Main Steam Flow event is conservative since the SLB event is a much more severe increase in heat removal event.
The last analysis of this event documents a
minimum DNBR of
- l. 63.
Accordingly, the expected decrease in minimum DNBR would result in a value well above the SAFDL of 1.28.
Inadvertent 0 enin of a
S.
G.
Safet Valve or Atmos heric Dum Valve The impact of an extra 0.4 second delay on a trip occurring at 830.3 seconds into the transient is negligible with respect to the two hour site boundary dose calculated.
0
2.0 Decrease in Heat Removal b the Seconda S stem 2.1 Loss of Condenser Vacuum The Loss of Condenser Vacuum (LOCV) event is the most limiting event in this category since it produces the highest calculated (peak)
RCS and secondary pressure.
The LOCV event was re-analyzed to demonstrate that with an increased CEDM holding coil
- time, the RCS and secondary pressures do not exceed the acceptance criteria of 2750 psia and 1100
- psia, respectively.
With an increased holding coil delay time of 0.74 seconds, it takes a longer time before the control rods begin to drop increasing the duration of the power excursion and consequently resulting in higher RCS and secondary pressures.
Attachment 4 provides a thorough discussion of the analysis, the assumptions and results following the UFSAR format.
The results of the re-analysis demonstrate that the LOCV event with an increased CEDM holding coil time will not result in peak RCS pressure or peak secondary pressure in excess of their respective upset limits.
The increased CEDM holding coil delay time results in an incremental increase in the calculated peak RCS and secondary pressure of 18 psia and 2
- psia, respectively.
Therefore, it can be concluded that the
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safety analysis criteria for the LOCV or any other DBE which 'challenges the pressure upset limits.are not violated with the proposed amendment.
2.2
~
I Loss of Normal A.C. Power The 0.4 second increase in CEDM holding coil delay has no impact on the results of this analysis since the trip occurs at 1.91 seconds (almost instantaneously) and the criterion is the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site dose release.
- Hence, the extra 0.4 seconds of CEDM holding coil delay is judged to have a negligible impact on the dose calculation results.
2.3 Feedwater S stem Pi e Breaks The effect of the increased CEDM holding coil delay is to increase the peak RCS pressure attained during a
Feedwater Line Break (FLB) event.
For the LOCV event, the peak pressure increased 18 psia for a 0.4 second increase in the holding coil delay.
Since the LOCV event, shows a more severe pressurization response with respect to peak RCS pressure than that of a FLB event, it is expected that the peak pressure increase observed for the LOCV event as a result of the increased CDEM holding coil delay will bound the expected impact on the results of
l'ga
the FLB event.
Therefore, assuming an RCS pressure increase of 18 psia for the peak RCS pressure calculated 4I for the FLB event is clearly conservative.
The analysis of record results show a peak RCS pressure of 2715 psia, therefore enough margin is available to accommodate this pressure increase.
The resultant peak RCS pressure expected with the proposed amendment is shown to be below the acceptance criterion of 2750 psia.
3.0 Decrease in Reactor Coolant Flowrate 3.1 Loss of Flow Protection against a
4-pump Loss of Flow (LOF) is provided by building sufficient thermal margin into the DNB related Limiting Conditions for Operation (LCO),
i.e.,
by including sufficient Reguired Overpower Margin (ROPM) and by the action of the Low Flow trip.
The ROPM conservatively accounts for the margin degradation from the initiation of a LOF and the time of minimum DNBR.
The LOF event is the limiting event with respect to establishing the LCO for protection against DNB.
The LOF event was re-analyzed to determine the minimum initial margin that must be maintained by the LCO's to ensure that the DNBR will not violate the SAFDL of 1.28 10
when an increased CEDM holding coil delay time of 0.74 seconds is assumed.
Attachment 4 provides a thorough discussion of the
- analysis, assumptions and results, following the UFSAR format.
A 3.5% decrease in overpower margin (or equivalently by a
3.5~
increase in ROPM) was calculated for the 0.4 second increase in CEA drop time.
This 3.54 margin degradation was checked with respect to the existing margin available.
Since there existed at. least 54 overpower margin between the actual calculated DNB LCO and the Technical Specification LCO, the 3.5w reduction can be accommodated without changing the existing DNB LCO in the Technical Specification.
3.2 Sheared Shaft The Single Sheared Shaft (SSS) event is the limiting event with respect to the rate of DNBR degradation.
The analysis of record for the SSS event predicted a minimum DNBR of 1.256.
At. that time, the value was reported to be 1.23 in order to obtain a conservative prediction of the fraction of fuel failure following a SSS event.
An evaluation was performed to determine the impact on the predicted minimum DNBR and the related fraction of failed fuel due to an increased CEDM holding coil delay time.
11
The calculated minimum DNBR for this event occurs at 1.3 seconds.
This time is 0.6 seconds after the trip breakers open and 0.26 seconds after the rods begin to fall into the core.
By this
- time, the RCS flow degradation has been completed and RCS flow remains constant for the rest of the transient.
Therefore, the impact on minimum DNBR due to an increased CEDM holding coil delay time is due mainly to the power and temperature effects and not to RCS flow degradation which typically has the more limiting effect on minimum DNBR degradation.
Results of the evaluation show that increasing the CEDM holding coil delay time decreased the predicted minimum DNBR by.34, to a new value of 1.252.
The calculated minimum DNBR is still bounded by the previously reported value of 1.23 for Cycle 2 therefore, it can be concluded that the results of the single sheared shaft event after implementation of the proposed amendment are bounded by the results presented in the analysis of record.
4.0 Reactivit and Power Distribution Anomalies 4.1 CEA E'ection from Hot Full Power The CEA Ejection Event from Hot Full Power (HFP) is the most limiting with respect to the limits on the deposited 12
A
energy in the fuel.
The design criteria used to determine fuel pin failure occurrence are:
1) clad damage total average enthalpy
=
200 cal/gm, and 2) incipient centerline melting total centerline enthalpy
= 250 cal/gm.
The CEA Ejection event was re-analyzed to evaluate the impact of an increased CEDM holding coil delay time of 0.74 seconds on the predicted deposited energy in the fuel.
In addition to the holding coil delay time analysis assumption, two other analysis assumptions were changed when compared to the existing analysis of record:
1)
The post-ejected radial peaking factor was reduced from 3.5 to 3.2 to reflect a value more characteristic of the actual calculated values for recent
- cycles, and 2) the assumed scram worth was conservatively reduced from -4. 54 hP to -3. 04 69 to accommodate expected reductions in available scram worths for future cycles.
With these two additional changes, the peak average and centerline enthalpy calculated for the hottest pellet were both below all the fuel deposited energy limits, therefore, no fuel failure is predicted to occur as a
result of the proposed amendment.
Attachment 4 provides a thorough discussion of the analysis, the assumptions, and results, following the UFSAR format.
13
k AQ)
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I
4.2 CEA E'ection Initiated from Zero Power The full power CEA ejection event re-analysis shows that the increase in CEDM holding coil delay time resulted in an 11 cal/gm average enthalpy rise and a
19 cal/gm centerline enthalpy rise.
The previous analysis of the CEA ejection event from zero power showed a total average enthalpy of 124 cal/gm and a total centerline enthalpy of 180 cal/gm.
- Hence, the expected increase in the two values due to the increased CEDM holding coil delay would result in a total average enthalpy of 135 cal/gm and a
total centerline enthalpy of 199 cal/gm; both of which are well below the acceptance criteria for the event.
4.3 Uncontrolled CEA Withdrawal Initiated from Zero Power Increasing the CEDM holding coil delay time willdecrease the previously calculated minimum DNBR of 2.55.
As was demonstrated in the Pre-Trip SLB event evaluations, the change in DNBR calculated to account for the increased CEDM holding coil delay is small, about 3>os
- Clearly, a
change of this magnitude can be accommodated for the CEA Withdrawal event from zero power without violation of the SAFDL.
14
0
~ J Jr
The LHR SAFDL will not be violated, since the LHR calculated was equivalent to a final centerline enthalpy of 180 cal/gm.
The re-analysis of the CEA ejection event from hot full power shows that the increased CEDM holding coil increases the deposited energy in the centerline at the most by 19 cal/gm; for a total of 199 cal/gm.
Clearly, this change can be accommodated in the CEA Withdrawal (CEAW) event from zero power without violating either the average deposited energy criterion of 200 cal/gm or the incipient centerline melt threshold of 250 cal/gm.
4 '
Uncontrolled CEA Withdrawal Initiated from Full Power It is judged that the RCS pressure increase will be less than in the LOCV event which was 18 psia.
Assuming an 18 psia increase in the CEAW HFP
- case, the peak RCS pressure attained would be 2550 psia plus 18 psia which clearly is well below 2750 psia.
For the limiting CEAW, the analysis of record calculated peak LHR was 16.38 kw/ft.
The increased CEDM holding coil delay time would allow the power excursion to continue for an additional 0.4 seconds.
The previous peak linear heat rate calculation for Full Power was based on a 204 fractional power rise; having a rate of 15
Nil t.
t,4
about 54 per second.
For this limiting case, a
0.4 second increase in holding coil delay would equate to a 24 increase in the maximum power change.
Thus, the new peak linear heat rate is about 16.7 kw/ft which is still well below the SAFDL of 22 kw/ft.
This event is also evaluated to determine the TM/LP setpoints.
Since the TM/LP trip is designed to protect the core against very slow power excursions where large amounts of CEA motion are required to produce the power excursion, an increase of 0.4 seconds in CEDM holding coil delay time willnot significantly impact the results of this analysis with respect to minimum DNBR considerations.
Therefore, it can'e concluded that the current TM/LP setpoints remain valid.
4.5
~CEA Dro The CEA drop event does not require a
scram to demonstrate adequate protection.
- Instead, sufficient margin is built into the Technical Specification LCO to accommodate the margin decrease due to a CEA drop event without requiring a trip.
Thus the CEDM holding coil delay time assumed has no impact on the results of this event.
16
4.6 Inadvertent Boron Dilution The boron dilution methodology does not depend on a trip to demonstrate adequate protection.
Accordingly, the increase in the CEDM holding coil delay time does not impact the results for this event.
5.0 Decrease in Reactor Coolant S stem Inventor 5.1 Pressurizer Pressure Decrease Events The pressurizer pressure decrease event is analyzed primarily to generate a pressure bias incorporated in the g term of the TM/LP trip equation.
The previous analysis produced aW bias of 41 psia.
Increasing the CEDM holding coil delay time by 0.4 seconds would delay rod insertion by 0.4 seconds.
This would be expected to move out the time of minimum DNBR by an equivalent amount.
Based on the rate of pressure change close to the time of minimum DNBR, an additional pressure decrease of 15 psia is conservatively predicted.
Therefore, the total Y bias calculated for this event when the CEA holding coil delay is 0.74 seconds is 56 psia.
This value is still below the 70 psia which was 17
0 41 I
tg I
Attachment 3
Determination of No Si nificant Hazards Consideration 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 10CFR50.92, 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 change does not affect any active hardware involving plant operation; rather it affects an acceptance criterion for confirming the required performance of the existing control element assembly (CEA) hardware.
Therefore, the proposed
change does not increase the probability of an accident previously analyzed.
The impact of changing the CEA drop time from 2.7 to 3.1 seconds on all safety analysis related Design Basis Events (DBE's), for which a
scram of the CEA's is predicted, was assessed by specifically re-analyzing only the most limiting events with respect to the various safety analysis fuel and system criteria.
In particular, the following events were re-analyzed:
Loss of Condenser Vacuum (LOCV)
Loss of Forced Reactor Coolant Flow Pre-Trip Steam Line Break (SLB)
Hot full Power CEA Ejection (CEA Ejection)
It has been demonstrated that the events are either totally unrelated to CEA drop time considerations or are not significantly impacted.
Additionally, it was demonstrated for each potentially impacted analysis that the consequences of the analysis remain unchanged or are bounded by the existing analysis.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Use of the modified specification would not create the possibility of a new or different kind of accident from any 1
accident previously evaluated.
The proposed change in the Technical Specifications does not affect any active hardware involving plant operation; rather it affects only an acceptance criterion for confirming the required performance of the existing CEA hardware.
Therefore, the proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.
(3)
Use of the modified specification would not involve significant reduction in a margin of safety.
The increased CEA drop time has been evaluated for its impact on the current licensed safety analysis.
The results of the re-analysis for those transients which are potentially impacted by the proposed change show that the reference analyses are valid or that the new analysis results still show acceptable results with respect to the acceptance criteria.
Therefore, there is no significant reduction in the margin of safety.
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
Regnal zed Events in U dated Final Safet An sis Re ort Format
Inside Containment Pre-Tri Power Excursions
~ ~
EI I
U A45
3.2.1.5a R
Failures in the main steam system piping were re-analyzed to ensure that a
eoolable geometry 1s ma1ntained and that s1te boundary doses do not exceed 10CFRI00 guidelines when the CEON holding co11 delay is 1ncreased from
.34 seconds to.74 seconds.
A rupture in the main steam system pip1ng increases steam flow from the steam generators.
This increase 1n steam flow increases the rate of RCS heat removal by the steam generators and causes a
decrease in core inlet temperature, In the presence of a negative moderator temperature coefficient of react1v1ty (NTC), the decrease in core inlet temperature causes core power to increase.
This 1ncrease in core power can cause the SAFDL's to be exceeded.
The excursion in core power is terminated by the action of one of the following Reacto~ Protective System (RPS) trips:
Thermal Margin/Low Pressure
{TH/LP), Low Steam Generator Pressure (LSGP),
H1gh Power or High Containment Pressure.
3.2. 1.5a.2 Anal sis of E Steam Line Breaks (SLB's) 1nside containment may be postulated to have break areas up to the cross section of the largest main steam pipe (6.305 ft ).
In evaluating SLB's inside the containment, environmehtal degradation of the sensors input to the dT-power calculator and the pressure measurement sensors is considered.
Also considered is the decalibration of the ex-core neutron detectors caused by the cooler water in the downcomer region.
SLB's with severely negative HTC's and large break
- areas, while initially causing a more severe power excursion, are mitigated earlier by action of RPS trips.
The earl1er act1on of the RPS trips is caused by:
1.
The more severe steam generator pressure
- decrease, accompanying SLB's with large break areas, causes an early LSGP trip.
2.
The 1ncreased mass and energy release due to the larger breaks into the containment atmosphere causes a
more rapid 1ncrease in containment pressure and earlier High Containment Pressure trip.
Consequently, it is necessary to perform the analysis parametric 1n break size and NTC to ensure that the limiting break area is identified.
Table 3.2.1,5a-l presents the initial conditions used in the analysis of the Inside Containment SLB Event based on these parametric evaluations.
A Loss oF AC Power (LOAC) fs postulated to accompany the SLB Event.
The LOAC causes a coastdown of the reactor coolant pumps.
Depending on when the LDAC occurs, it can result in a lower RCS flow during at the time period when the peak heat flux is present.
The LOAC was assumed to occur at a
time to maximize the calcu1ated potential fuel damage.
This time was determined to be when the LDAC would cause a Low Flow trip sfmultaneous wfth a High Containment Pressure trip.
3.2,1.5a.3 A parametric analysis fn both HTC and break area was performed on the Inside Containment SLS Event.
This parametric analysis identified the limiting Insfde Containment SLB Event in terms of maximum heat flux before an RPS trip; The 'limiting inside containment PB Event uas found to be the break cau]ing an effective flow area of 2.01 ft with an effective HTC of -.54xlD hp/'F.
Table 3.2.1,5a-2 and Figures 3.2.1.5a-l through 3.2.1.5a-6 present the sequence of events for the limiting Inside Containment SLB Event and present the response of the NSSS for power, heat flux, RCS temperature, RCS pressure, steam generator pressure and reactivity.
LI~ I I
~ I The ana'lysfs of the Inside Containment SLB Event demonstrates that a eoolable geometry is maintained as the number of fuel pins predicted to fail is less than 10 percent.
Site boundary doses for the inside containment SLB are bounded by the doses obtained in the outside containment SLB analyzed in Section 3.2.1.5b.
TABLE 3.2.1.5a-l Total RCS Power (Core Thermal Power including 24 power measurement uncertainty 5 pump heat)
Initial Core Coolant Inlet Temperature Initial Reactor Coolant System Pressure Initial Steam Generator Pressure CEA Morth at Trip Effective Moderator Temperature Coefficient 4 Fraction {Including Uncertainty}
CEO'old)ng Coil Delay Time psia psia
%hp x10 hp/'F sec 2774 552 2170 910
~7 e3
-2.7 to -.27
.0060
.74
NC TABLE 3.2. 1.5a-2 Xin~ml 0.0 33.2 33.4 37.10, 37.75 38.49 38.9 51.6 52.3 54.1 Failure in the Hain Steam System Piping LOAC on Turbine Trip, Reactor Coolant Pumps Begin to Coastdown Haximum Core Power RPS Trip Generated, Low Coolant Flow (High containment pressure trip would have occurred at this time without the LOAC assumptions.)
Tr)p Breakers Open CEA's Begin to Enter Core H)nimum DNBR Safety Infection Signal HSIS Generated on Low Steam Generator Pressura Pressurizer Empties 2.01 ft 136.1%
.?82 1580 psia'60 psia
140 120 I-
~I 80 UJL 60 40 20 20 40 60 SO
. TINE SECONDS FLORIDA POHFR 1
L IGHT COa St. hcfe i Nue1ear Power Plant SFN U% SK
%K'NSlK CSTAtNNÃ PRE-TfUP f6&FXOBSIyS NK RKR YS TIE HSJhg 3 2i1,%-1
iGQ 140 20 40 60 80 TINE e SECONDS FLORIDA
$'ONKR k LIGHT Cas St. 4c<e 2 NucI ear Paver P1ant Sly UNE HlBK&ST MIKQNTAIN%NT PK-TMP POLY B(GNINS KK %AT RlN TfK'
660 BOO outlet aYerage B40 V7 LQ CL S>O Q.
4J
~. 4SO
)nlet 450 20 40 60 80 T I NE i SECONDS 100 Ft.ORIANA POWER 4 L.IGHT CA<
St. Lucfe 2 Nuclear Power Plant S7EN LIhK BiBK BENT INSIEE QNTAINNPK-TRIP le& BQISIQ6.
teGTOR COOLNT SYSEl KN'ERATUEIES VS TI%
FIGSK 3,2)l)5A-3'
1950 1800 1650 13SO 1200 20 40 60 90 TIME.
SECONDS 100 PLOR IPA POWER
& LIGHT CO e St. Lucre g
Rlcltal'owot'lLhC S79N UNE BfBX BENT l%IKGFAINZ RK-TRIP RKR ENMIR PEAClM QCVNT SYS7EN PKRRK VS TI%
FISNE 9,2<1,SAN
)000 900
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Loss of Condenser Vacuum 3.2.2.3
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3.2.2.3 3.2,2.3.1 The Loss of Condenser Vacuum event was re-analyzed to demonstrate that the RCS and main steam system pressures do not exceed 110% of design values (i.e.,
2750 psia and 1100 psia, raspect1vely) when the CEDM holding co11 delay time assumed is 1ncrease from 0.34 seconds to 0.74 seconds.
The Loss of Condenser Vacuum (LOCV) w111 cause the turbine stop valves to
- close, the main feedwater pumps to trip, and the steam bypass valves to be unavailable.
No credit was assumed for a simultaneous reactor trip on turb1ne trip (dua to closure of turbine stop valves).
The result1ng loss of load and loss of main feedwater causa tha steam genera'tor pressure to increase and open
~
The transient 1s term1nated by a reactor trip on high pressurizer pressure.
3.2.2.3.2 The LOCV event was initiated at the conditions shown in Table 3.2.2.3-1.
The combination of parameters shown in th1s table maximizes the calculated peak RCS pressure.
The most important parameters for this event, other than HTC and the number of plugged tubes, are the initial primary inlet temperature and the fuel temperature coefficients of reactivity.
The methods used to analyze this event are consistent with those described 1n the FSAR.
The initial core average axial power distribution for this analysis was chosen to ba a
bottom peaked shape.
This distribution is assumed because it minimizes the negative reactivity inserted dur ing the initial portion of the scram following a reactor trip. 'his will delay reactor shutdown and hence max1mizes tha pressure transient.
A Moderator Temperature Coefficient (NTC) of 0.3xl0 hp/'F (the most positive value allowed at Hot Full Power (HFP)) was used in the analysis.
This HTC causes the greatest amount of positive reactivity feedback to exacerbate the transient increase in pressure.
A Fuel Temperature Coeff1cient (FTC}
corresponding to beginning of cycle conditions was used in the analysis.
This FTC causes the least amount of negative react1vity feedback to mitigate'he transient increase in pressure.
Tha uncertainty on the FTC used in the analysis is shown 1n Table 3.2.2.3-1.
A minimutti allowable 1nitial RCS pressure is used to maximize the rata of change oF pressure.
Th1s will maximize peak pressure following trip.
The lower RCS inlet temperature lowers the initial steam generator pressure resulting in a mora severe secondary transient due to the delay in opening of the main steam safety valves.
In addition, a
loss of offsita power was assumed to occur such that the high pressurizer pressure trip and tha low coolant flow trip occur simultaneously.
This max1mizes the peak RCS pressure following trip.
An assumption that 1500 tubes were plugged in each steam generator was made.
This decreases the primary-to-secondary heat transfer and therefore, increases the primary tamperaturas and also the maximum RCS pressure attained during the transient.
A holding coil delay time of 0,74 seconds was assumed.
This results in a
longer time before the scram rods begin to drop, which results in higher RCS and secondary pressures.
3.2,2.3.3 ~~y, The LOCV event, initiated from tha conditions given in Table 3,2.2.3-1, results in a high pressurizer pressure trip condition at 5.45 seconds.
At IO.O seconds,'he primary pressure reaches its maximum value which does not exceed the 2750 psia limit.
The 1ncrease in secondary pressure is limited by the opening of tha main steam safety valves, which open at 9.6 seconds.
The secondary pressure reaches its maximum value which does not exceed the ll00 psia limit at 17. I seconds, after the initiation of the event.
Table 3.2.2,3-2 presents the sequence of events for this event.
Figures 3.2.2.3-1 to 3.2.2.3-5 show the NSSS response to power, heat flux, the RCS
- pressure, RCS coolant temperatures, and steam generator pressure.
The core performance following a
LOCV would be no more adverse than those following a-Loss of Normal AC Power, wh1ch is described in Section 3.2.2.4, The radiolog1cal consequences due to steam releases from the secondary system would ba lass severe than the consequences of the Inadvertent Opaning of an Atmospheric Oump Valve Event discussed in Section 3.2.1.4.
3.2.2.3.4 The results of this re-analysis demonstrates that the Loss of Condenser Vacuum event with the longer CEDH holding coil delay time will not result in peak RCS ressura or ma1n steam pressure in excess of their respective upset pressure imits.
ARAM E
F N
Emmaiaz Total RCS Power (Core Thermal Power
+ Pump Heat)
Initial Core Coolant inlet Temperature Initial Reactor Coolant System Pressure Initial RCS Vesse1 flow Rate Moderator Temperature Coefficient Ooppler Coefficient Multiplier CEA North at Trip Number of Plugged Tubes GEOM Holding Coil Oelay Time
'F psia gpm x10 dp/'F 0/S.G.
sec 2774 535 2170 363,000 0.3
.85
-5.5 1500
.74
ORT 0
0.0 4.6 6.45 6.60 7.34 7.50 10.0 13.9 17,1 Closure of Turbine Stop Valves on Turbine Trip due to Loss of Condenser Vacuum Loss of Offsite Power High Pressurizer Pressure Trip/
Low Flow Trip Analysis Setpoint Reached Pressurizer Safety Valves Open Trip Breakers Open CEAs Begin to Drop Into Core'aximum Core Power Steam Generator Safety Valves Open Maximum RCS Pressure*
Pressurizer Safety Valves Close Total PSV Release Maximum Steam Generator Pressure 2428 psia/93Fi of 363,000 gpm 2525 psia
<107.45 of 2700 HWt 1010 psia
<2750 psia 2424 psia 1216 ibm
<1100 psia
+ Including pump and elevation head
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Total Loss of Forced Reactor Coolant Flow 3.2.3.2
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3 2.3.2 3.2.3.2.1 The Loss of Coolant Flow (LOF) Event was re-analyzed to determine the m1nimum initial margin that must be maintained by the Lfmit1ng Cond1tfons for Operations (LCOs) such that 1n conjunction with the Reactor Protection System (RPS),
the DNBR wf11 not be less than the SAFDL of 1.28 when the CEOH holding coil delay time assumed fs increased from.34 second to.74 second.
A loss of normal coo1ant flow may result either from a loss of electrfcal power to one or more of the four reactor coolant pumps or from a mechanical
- Fa11ure, such as a pump shaft seizure.
Sfmultaneoqs mechanical Fa1lure of two or more pumps is not cons1dered credible.
Under voltage or under frequency of the motor drive electrical power source can result fn a reduction of coo'lant flow; and, 1f the flow reduction from either cause is greater than the Low Flaw trip setpoint, a reactor trip fs initiated.
The design of the ons'fte electrical power system for the reactor coolant pumps fs such that no single failure From a normal operating condition can cause a
complete 4-pump LOF Event where the pump-motor-flywheel combination 1s the only available source of coastdown energy.. Nevertheless, for the purpose of this accident evaluation, a complete 4-pump LOF is conservatively assumed.
The 4-Pump LOF Event produces a rap1d approach to the DNBR limit due to the rapid decrease in the core coolant Flow.
Protection against exceeding the DNBR limit for this transient is provided by the initial steady state thermal margin which fs maintained by adhering to the Technical Specifications'CO's on DNBR margfn and by the response of the RPS.
Reactor trip on loss of coolant flow 1s tnit1ated by a low coolant flow rate as determined by a reduction fn the sum of the steam generator hot to cold leg pressure drops.
This signal fs compared with a setpofnt which is a Function of the number of energized reactor coolant
- pumps, For a loss of flow at full power operating condition, a trip will be 1nftiated when the flow rate drops to 93 percent oF rated flaw.
3.2.3.2.2 The transient fs characterized by the flow coastdown curve given fn Figure 3,2.3.2-1.
Table 3.2.3.2-1 presents the initial conditions assumed in this event.
The event fs analyzed parametric on axial shape
- index, to determine the maximum required over-power margin needed to ensure the SAFDL's are not violated.
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3.2.3.2,3 )hag~
Table 3.2,3.2-2 presents the sequence of events for the 4-pump Loss of Flow Event initiated at a
shape index of -.2.
The Low Flow tHp setpoint is reached at
.71 seconds and the scram rods start dropping into the core 1.40 seconds later.
A minimum CE-1 DNBR of 1.28 is reached at 2.4 seconds.
Figures 3.2.3.2-2 to 3.2.3.2-5 present the core
- power, heat
- flux, RCS
- pressure, and core coolant temperatures as a
function of time.
Figure 3.2.3.2-S presents the hot channel ONBR vs.
time for the case that is characterized by an axial shape index of -.20.
3.2.3.2.4 A 3.5%.increase in ONB margin degradation was observed when the longer CEOH holding coil delay time was util)@ed.
- However, there existed sufficient margin to show that the event initiated from the Technical Specification LCO's in con)unction with the Low Flow trip will not exceed the design DNBR limit.
This means that the minimum
- DNBR, of 1.28 reached in past
- analysis, is achieved with 3.55 less margin to spare than before.
S M
TABLE 3.2.3.2-1 V
Total RCS Power (Core Thermal Power
+ Pump Heat)
Initial Core Coolant
?nlet Temperature
..Initial RCS Vessel Flow Rate
'nit1al Reactor Coolant System Pressure Moderator Temperature Coefficient Doppler Coefficient Multiplier Low Flow Trip Response Time CEDM Holding Coil Delay CEA Time to 905 Insertion (Including Holding Co11 Oelay)
CEA Worth at Trip (all rods out)
Total Unrqdded Radial Peak1ng Factor (F) 4-Pump RCS Flow Coastdown
'F ppm psia xl0 hp/'F sec sec sec 2720+
549+
377,500 22254
~ 85 0.65 0.74
'.7
-7.0 F1gure 3.2,3.2-1
+
For ONBR calculations, effects of uncertainties on these paramet'ers were comb1ned statistically,
TABLE 3.2.3.2-2 0.0 Loss of Power to all Four Reactor Coolant Pumps
.71 1.36 2.10 Low Flow Trip Signal Generated Trip Breakers Open CEAs Begin to Drop into Core 93K of Rated Flow 2.4 Hinimum CE-1 DNBR
>1,28
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Control Element Assembl E'ection 3.2.4.6
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3.2.4.6 3.2.4.6.1 The Control Element Assembly (CEA) Kgectfon Event fs re-analyzed to determfne the fract1on of fuel pfns that exceed the criteria for incipient centerline melting and rad1al average enthalpy when the CEDM hold1ng coil delay assumed fs increased from.34 seconds to.74 seconds.
Rapid e)ectfon of a CEA from the core would require a complete circumferential break of the control element drive mechanism (CEDN) housing or of the CEDH nozzle on the reactor vessel head.
The CEDE housing and CEDM nozzle are an extension of the reactor coolant system boundary and designed and manufactured to Section III of the ASNE Boiler and Pressure Vessel Code.
- Hence, the occurrence of such a failure fs considered highly unlikely.
A typical CEA e]ectfon transient behaves fn the following manner:
After edectfon of a
CEA from a
Hot Full Power (HFP) fnit1al condition, the core power rises rapidly For a brief period, The rise fs terminated by the Doppler effect.
Reactor shutdown fs initiated by the high power level
- trip, terminating the remainder of the power transient.
The core is protected against severe fuel damage by the CEA insertions permitted at various power levels by the Power Dependent Insertion Limit (PDIL) Technical Specification and by the high power trip.
3,2.4.6.2 The key parameters used 1n this event are listed in Table 3.2.4.6-1.
fifth these key parameters, selected to add conservatism, the average and centerline enthalpfes 1n the hottest spot of the fuel rod are calculated.
The calculated enthalpy values are compared to threshold enthalpy values to determine the amount of fuel exceeding these thresholds.
These threshold enthalpy values are:
Clad Damage Threshold Total Average Enthalpy
~ 200 cal/gm Incipient Centerline Melting Threshold Total Centerline Enthalpy 250 cal/gm Fully molten Centerline Threshold Total Centerline Enthalpy
~ 310 cal/gm
Fuel exceeding these thresholds are assumed to fail for the ourpo'se of establishing site boundary doses.
To bound the most adverse condit)ons during the cycle, the most limiting of either the Beginning of Cycle (BOC) or End of Cycle (EOC) values of key parameters were used in the analysis.
A BOC Doppler defect was used since it produces the least amount of negative reactivity feedback to mitigate the tr'ansient.
The Technical Specification most positive moderator temperature coefficient of +0,5xl0 dp/'F was used because a
positive HTC results in positive reactivity feedback and thus increases the magnitude of the power rise.
An EOC delayed neutron fraction was used in the analysis to produce the highest power rise during the event.
The full power case was
- ana1yzed, assuming the value of 0.0S seconds for'he total ejection time, which is consistent with the FSAR.
No credit was taken for voids.
3.2.4.6.3 Rag~
The power transient produced by a CEA exsection initiated at the HFP is shown on Figure 3.2,4.6-1.
The results of the CEA egection case analyzed (Table 3.2.4.6-2) show that the maximum total energy deposited during the event is less than the criteria for both clad damage (i,e.,
200 cal/gm) and incipient center line melting (i.e.,
2N cal/gm).
Consequently, no fuel pin failures are predicted to occur.
3.2.4,6.4 Since no failure of fuel is predicted to occur, site boundary doses will be limited to a small fraction of 10GFR100 guidelines and hence the consequences of this event are acceptable.
TABLK 3.2.4.6-1 ON KQLPsum Total RCS Power (Core Thermal Power + Pump Heat)
Core Average Lfnear Heat Generatfon Rate at 2774 IOt moderator Temperature Coefficient
@ected CEA North Delayed Neutron Fraction Post-E)ected Radial Power Peak Axial Power Peak CPA Worth at Trfp Doppler Coefffcfent Hultfplfer CEOH Holding Coil Delay Time kH/ft x 10 hp/'F
%hp sec 2774 4.81 t.S
.25
,0044 3.2 1.64
-3.0 0.85 0.74
TABLE 3.2 4,6-2 QQ,LEmae Total Average Enthalpy of Hottest Fuel Pallet (cal/Qm)
Total Centerline Enthalpy of Hottest Fuel Pe11et (cal/gm)
Fraction of Rods that Suffer Clad Oamage (Average Enthalpy >200 cal/gm)
Fraction of Fuel Having at Least Incipient Cinterline Melting (Center 1ine Kntha1py
>260 cal/gm)
Fraction of Fuel Having a Fully Molten Centerline Condition (Centerline Enthalpy >310 cal/gm)
<200
<250
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