ML20137L016
| ML20137L016 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/02/1997 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137K821 | List:
|
| References | |
| FOIA-96-485 JPN-PSL-SENP-95, JPN-PSL-SENP-95-101, NUDOCS 9704070139 | |
| Download: ML20137L016 (16) | |
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FLORIDA POWER & LIGHT CO 4
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( ENGINEERING EVALUATION i
Assessment of the Effects on Plant Operation of Lifting the LPSI Pump Discharge Header Thermal Relief Valve 4
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BT LUCIE NUCLEAR PLANT i
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UNIT 1 i
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9704070139 970402 i
a JPN-PSL-SENP-95-101 REVISION 2 PAGE 2 OF 9 REVIEW AND APPROVAL RECORD l
PLANT ST. LUCIE UNIT 1
TITLE Assessment of the Effects on Plant Operation of Liftina the LPSI Pump Discharae Header Thermal Relief Valve i
LEAD DISCIPLINE LICENSING ENGINEERING ORGANIZATION PR00VCTION ENGINEERING GROUP REVIEW / APPROVAL:
INTERFACE TYPE GROUP PREPARED VERIFIED IPPROVED FPL APPROVED *
[NPUT REVIEV N/A mmm MECH X
ELECT X
!&C X
CIVIL X
NUC FUEL X
PSA X
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- For contractor Evals As Determined By Projects
- Review Interface As A Min On All 10CFR50.59 Evals and PLAs FPL PROJECTS APPROVAL:
DATE:
OTHER INTERFACES None JPN Form 24, Rev. 9/92
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- l-JPN-PSL-SENP-95-101 REVISION 2 i
PAGE 3 OF 9 TABLE OF CONTENTS
!I 1
i SECTION TITLE 4
PAGE Cover l
1 4
Review cnd Approval Record 2
3 Table of Contents 3
1.0 Purpose and Scope
4
2.0 Background
t 4
3.0 Design Bases 4
)
4.0 Analyses of the Event 5
5.0 Conclusions 9
j 6.0 Verification Summary 9
7.0 References 9
Attachments None
o JPN-PSL-SENP-95-101 j
REVISION 2 PAGE 4 OF 9 1.0 PURPOSE AND SCOPE On August 10, 1995, low pressure safety injection pump (LPSI) discharge header thermal relief val're V3439 lifted releasing about 4000 gallons of reactor coolant into the reactor auxiliary building (RAB).
The purpose of tnis evaluation is to assess the significance of this event on plant operation.
More specifically, this evaluation assesses the effects of this event during normal, shutdown and design basis accident conditions.
Revision 2 to this evaluation provides more detail and discussion on the assumptions and the qualitative assessment of the radiological consequences for the selected events.
The original conclusions of this evaluation remains unchanged.
This engineering evaluation involves engineered safeguards systems and is therefore classified as safety related.
2.0 BACKGROUND
On August 10, 1995, reactor coolant water was discovered flowing from the pipe tunnel door at the -0.5 elevation of the RAB (Ref 1).
Subsequent investigation revealed that the LPSI pump discharge header thormal relief valva (V3439) was lifting and would not reseat.
Based on the dimensions of the pipe tunnel and depth of water, and the amount of water that filled the aerated waste storage tank, it was estimated that approximately 4000 gallons of water had discharged from the relief valve.
In addition, the safeguards pump room sump isolation valves (which includes the pipe tunnel drain valves) were found to be closed and the two western pipe tunnel floor drain lines were clogged.
The relief valve flow rate is 40 gpm at the relief setpoint.
3.0 DESIGN BASIS REVIEW LPSI Discharae Header Thermal Relief Valve: (FSAR Section 6.3.2.2.6.c)
The LPSI discharge header thermal relief valve provides overpressure protection due to fluid thermal expansion for an isolatable section of the LPSI system discharge piping.
RAB Waste Manacement System The RAB waste management system includes the pipe tunnel floor and drainage system which is the discharge point for the thermal relief valve.
This discharge is directed via the RAB drain system to the safeguards room sumps.
Water in this sump is normally pumped automatically into the equipment drain tank (EDT) upon high sump level where it is then pumped into the aerated waste storage tank (AWST).
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1 I
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JPN-PSL-SENP-95-101 REVISION 2 PAGE 5 OF 9 From the. control room, an alternate path can be aligned to pump the safeguards sump water to the reactor drain tank (RDT) in the containment building (FSAR Section 11.2.2.1).
This alternate path is the leakage collection and return system (LCRS).
It is a dedicated _ system which is normally aligned to the reactor drain tank folicying a recirculation actuation signal (Ref 2).
The."A" and "B" safeguari: room, sumps each have 2 sump pumps, each with a design operating flow rate of 50 gpm at 100 ft. of head.
The sump pumps are powered via non-vital MCCs 1A2/1B2.
The safeguards high and high-high sump alarms are powered from vital MCC 1AB.
The function of the safeguards pump room sump isolation valves (HCV-25-1 through 7) is to isolate the safeguards rooms from RAB flooding due to a fire main break (FSAR Section 9.5A-110).
The safeguards pump room sump isolation valves are normally open to allow drainage to the ECCS sump.
o 4.0 ANALYSIS OF EFFECTS OF LIFTING V3439 The lifting of V3439 will occur while on SDC'if LPSI pump operation results in LPSI discharge. header pressures reaching the relief valve setpoint (500 psig i 3%).
Sufficient pressure to lift the relief valve may occur during LPSI pump starts that result in a j
pressure spike (=30 psi).
The pressure spike in conjunction with the pumps AP of 182 psi and a suction pressure of between 257 and 268 psia, will reach the minimum design setpoint.
With the blowdown for the valve set at 14%, the valve will continue to relieve until system pressure drops below approximately~430 psig.
Therefore, the analysis ' in this evaluation will only consider operation that involves LPSI discharge pressures approaching the relief valve setpoint.
(LPSI discharge pressures will not exceed relief valve setpoint during operation with the shutdown cooling system isolated from the RCS.)
Loss of RC8 Inventory If V3439 lifts during normal shutdown cooling (SDC) operation, loss of reactor coolant will occur until SDC is secured or the relief valve reseats.
Since the flow rate through the valve at the setpoint pressure is limited to about 40 gpm, the rate of inventory loss is well within normal charging system capability.
During long term cooling following a design basis accident the SDC system may be used to satisfy the RCS heat removal safety function.
Assuming the SDC system is being utilized, then the inventory loss of 40 gpm is well within makeup capability of either the high pressure safety injections pumps or charging pumps.
JPN-PSL-SENP-95-101 REVISION 2 PAGE 6 OF 9 Loss of RWT Inventory i
During the injection phase of small break loss of coolant accidents (LOCAs), excess steam demand events (ESDEs) and steam generator tube rupture (SGTRs) with LPSI pumps align' d to the refueling water e
tank (RWT), LPSI pump maximum discharge pressure is about 230 psig (RWT head plus LPSI pump Ap).
This is well below the lift setpoint of V3439.
Therefore a loss of RWT inventory would not occur.
Loss of containment Sume Inventory 4
During the recirculation phase of LOCAs, the LPSI discharge header pressure will not approach the lift setpoint of V3439.
This is because the LPSI pumps are either automatically stopped on receipt i
of a recirculation actuation signal, or if used for hot leg injection, maximum LPSI discharge header pressure would be = 260 i
j psia (containment sump head plus LPSI pump Ap), well below the lift setpoint of V3439.
Therefore a loss of containment sump inventory would not occur.
Floodina in the Pire Tunnel During the subject event the pipe tunnel floor drains were isolated which caused flooding in the pipe tunnel up to the height of the lip on the door (210").
Reactor coolant then spilled into the RAB hallway where it was then routed to the "B" safeguards room sump via an RAB floor drain.
The lower levels of the pipe tunnel do not contain electrical equipment that is susceptible to faults from a flooding event of this type.
When the safeguards pump room sump isolation valves are open, the discharge of V3439 is directed to the floor drains in the RAB pipe tunnel at the
.5 ft. elevation.
These drains empty into the safeguards room sumps via normally opened air operated valves controlled from a single switch on RTGB 106. The capacity of these drains exceeds the 40 gpm relief discharge, and therefore, the pipe tunnel will not flood if the drains are available.
Ploodinc in the "B" Safeguards PumD Room The "B" safeguards room sump holds 1100 gallons and automatically pumps down to the equipment drain tank (EDT) at c 850 gallons at a rate of 50 gpm (capacity of one sump pump).
If level reaches 950 gallons, the second sump pump starts.
Assuming non-vital power to the pumps is maintained, flooding of the "B" safeguards room will not occur.
Assuming a loss of offsite power (i.e., no safeguards sump pumps available), and that the discharged water is undetected (i.e., no operator intervention), the water level in the safeguards room would reach the "B"
HPSI pump motor in
=
7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
Annunciation of the safeguards high and high-high sump alarms would require investigation to determine the reason for.the alarm (Ref 3).
Seven hours is considered sufficient time to identify and isolate the leak prior to loss of the "B" HPSI pump.
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l JPN-PSL-SENP-95-101 REVISION'2 PAGE 7 OF 9 i
Assessment of Radiolocical Consecuences i
An assessment, as documented
- herein, was performed of the
[
radiological consequences of lifting V3439 concurrent with postulated FSAR Chapter 15 category accidents.
The purpose of the j
assessment was to determine the safety significance of the release 4
of reactor coolant into the RAB from the lifting of V3439 during shutdown cooling operation on 8/10/95.
The FSAR was first reviewed to determine which accident categories would be susceptible to the lifting of V3439.
Only those accidents which involved at least a partially pressurized RCS were selected since those events would have a sufficiently elevated RCS pressure to cause the relief valve to lift when going on shutdown cooling.
For this review, excess steam demand events (ESDEs), small break
' loss of cooling accidents (SBLOCAs) and steam generator tube rupture events (SGTREs) were considered bounding for other events in the FSAR that may involve shutdown cooling operation at a sufficiently high RCS pressure to cause V3439 to lift.
The assessment next reviewed the FSAR assumptions for the selected events and identified those assumptions that were not considered either reasonable or credible, or consistent with the operating history of the plant.
A revised set of reasonable and credible assumptions were then applied to a qualitative assessment of the radiological consequences for the selected events as described below:
ESDEs FSAR Dose Analysis ESDEs are characterized by rapid and significant cooling of the RCS due the faulted S/G.
In these events, the most reactive control rod is normally assumed to be stuck out of the core.
The positive reactivity-addition from the stuck rod and the cooler RCS
+
temperatures causes the reactor to experience an overpower transient which results in fuel exceeding minimum DNBR and failed fuel limited to $ 1.6% of the core.
The FSAR doses for this event result primarily from these fuel failures.
Revise,d Dose Assessment with V3439 Lifting St. Lucie Plant has never had a failure of its control rods to insert upon reactor trip from power operation.
With all rods inserted during an ESDE, the reactor would neither return to power,
' nor would it experience excessive power distribution anomalies, and would experience negligible if any fuel failure.
Therefore, the lifting of V3439 following this event would not involve a significant. release of radioactivity, and the FSAR dose analysis for this event remains bounding when compared to the results of this revised dose assessment (i.e., no fuel failures).
l JPN-PSL-SENP-95-101 i
REVISION 2 PAGE 8 OF 9 SBLOCAs i
FSAR Dose Analysis The FSAR analyzed SBLOCA includes depressurization of the RCS, high pressure safety injection 'at 1275 psia, dumping of the safety j
injection tanks at 215 psia and' low pressure safety injection at 200 psia.
The FSAR analysis for this event concludes that the i
Large Break LOCA doses are limiting.
Although not documented in the FSAR, the percentage of failed fuel is estimated not to exceed i
it.
Offsite doses would be from this failed fuel.
Note that for j
this limiting SBLOCA if shutdown cooling operation were to occur it would be at a very low RCS pressure after securing high pressure i
safety injection, and therefore, would not involve the lifting of V3439.
i Revised Dose Assessment with V3439 Lifting Only very limited SBLOCAs, involving partial,depressurization of j
the RCS, provide the conditions necessary to lift V3439.
For very limited SBLOCAs, either high pressure safety injection or charging will provide sufficient makeup, and fuel failure, if any, would be extremely small.
Therefore, the lifting of V3439 following this event would not' involve the significant release of radioactivity, and the FSAR-dose analysis for this event remains bounding when compared to the results of this revised dose assessment (i.e., none or much less than 1% fuel failures).
i SGTREs 1
FSAR Dose' Analysis i
Since the SGTF.E involves shutdown cooling operation following a controlled reduction in RCS pressure, it could result in lifting of V3439.
The FSAR analysis of this event provides offsite doses resulting from plant operation at the Technical Specification i
allowed maximum limits'of D2Q 131 concentration coupled with a one gpm primary to secondary leakage.
Fuel failures do not occur as a result of this event.
Revised Dose Assessment with V3439 Lifting St. Lucie Plant has not operated tj th primary to secondary leakage at the Technical Specification limit nor with DEQ 131 concentrations at the Technical Specification limit.
Therefore, the lifting of V3439 following this event would not involve a significant release of radioactivity, and the FSAR dose analysis for this event remains bounding when compared to the results of this revised dose assessment (i.e.,
limited primary to secondary leakage and reactor coolant activity).
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4 JPN-PSL-SENP-95-101 REVISION'2 i
- PAGE 9 OF 9 i
'In
- summary, a
qualitative assessment of the-radiological consequences of a selected set of bounding events representing FSAR Chapter 15 accident categories, concurrent with the lifting of 7
V3439, was performed to determine the safety' significance of the 8/10/95 event involving the inadvertent lifting of V3439.
]
Reasonable and credible assumptions, consistent with the operating experience of the plant, were opplied in the assessment.
When comparing the results of the:< assessments to the FSAR doses for i
the selected accident ' categories, the consequences of the FSAR j
analyses remain bounding.
f
5.0 CONCLUSION
S The safety significance of lifting V3439 with respect to plant operation has been evaluated.
The effects of loss of inventory, RAB flooding and. radiological consequences were reviewed and it is j
concluded that the lifting of V3439 would not have a significant f
effect on safe plant operation during normal, shutdown and design basis accident conditions.
i 1
6.0 VERIFICATION
SUMMARY
The scope of this verification was to review the inputs to j
determine if the results were reasonable. The method used for this verificatien consisted of ensuring that the applicable references, j
codes, and regulatory requirements were identified and addressed.
The inputs are correctly selected and applied.
i The conclusions provided are reasonable with respect to the inputs and discussions.
The verifier concurs with the Nuclear Safety Related classification of this Engineering Evaluation.
The i
rationale in assigning the safety classification was verified against the requirements of JPN Quality Instructions. The verifier
}
concurs with the conclusions outlined above.
7.0 REFERENCES
1)
In-House Event IHE-95-046, dated 8/17/95.
2)
St Lucie Unit 1 Emergency Operating Procedure 1-EOP-03, Loss of Coolant Accident, Rev 12.
3)
St Lucie Unit 1 Plant Annunciator Summary Procedure ONP 1-0030131, Rev 60 (Window R-14) 4)
St Lucie Unit 1 FSAR, through Amendment 14.
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REVISION 0 i
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TITLE Assessment of the Effects on Plant Operation of Lifting the LPSI Ptano Discharoe Header Thermal Relie. Valvc i
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PAGE 3 OF S TABLE OF CONTENTS I
i SECTION TITLE PAGI 1
Cover l
Review and App' oval Record 2
i Table of Contents 3
j 1.0 Purpose and Description 4
j I
2.0 Background
4 1
3.0 Design Bases 4
4.0 Analyses of the Event 5
5.0 Conclusions 7
6.0 Verification Summary 7
7.O References S
Attachments l
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I JPN-PSL-SENP-95-101 REVISION 0 i
PAGE 4 OF 8 io r
1.0 701 POSE AMD SCOPE j
On August 10, 1995, low pressure safety injection pump (LPSI)
J discharge header thermal relief valve V3439 lifted releasing about j
4000 gallons of renctor coolant into the reactor auxiliary building l
l (ras).
this evaluation is to assess the signifJeance of The purpose of this event on plant operation. More specifically, this e":a.uation assesses the effects of this event during normal, shutdown and b
design basis accident conditions.
This engineering evaluation involves engineered safeguarus systems and is therefore classified as safety related.
2.0 BACKGROUND
reactor coolant water was discovered flowing On August 10, 1995, from the pipe tunnel door at the -0.5 elevation of the RAB (Ref 1).
Subsequent investigation revealed that the LPSI pump discharge i
relief valve (V34391 was lifting and would not header thermal
'BasMn the dimensions 'of tEs" pipe ~ tunnel ~and depth 'of rensat.
w'at f and the amount of water that filled the aerated waste storage tank, it was estimated that apprarimately 4000 gallons of from the relief valve.
In addition, the water had discharged safeguards pump room sump isolation valves (which includes the pipe tueel drain valves) were found to be closed and the two western The relief valve flow
.unnel floor drain lines were clogged.
rate is 40 gym at the relief setpoint.
3.0 Desian Basis aeview
(FSAR Section j
LPSI Discharce Header Thermal Relief Valve:
6.3.2.2.6.cl The LPSI discharge header thermal relief valve provides overpressure protection due to fluid thermal expansion for an isolatable section of the LPSI system discharge piping.
RAB Waste Manaamnent System The RAB waste management system includes the pipe tunnel floor and drainage system which is the discharge point for the thermal relief This discharge is directed via the R&B drain system to the Water'in this samp is normally pumped valve.
"B" safeguards room sump.
automatically into the equipment drain tank (EDT) upon high sump level where it is then pumped into the aerated waste storage tank (ANST).
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il-JPN-PSL-SENP-95-101 REVISION O PAGE 5 OF 8 I
From the control room, an alternate path can be aligned to pump the safeguards sump water to the reactor drain tank (RDT) in the This alternate path containment building (FSAR Section 11.2.2.1).
and return system (LCRS).
It is a is the leakage collection is normally aligned to the reactor drain dedicated system whichfollowing a re W culation actuation signal (Ref 2).
tank (RDT) l e-'in, The "A" and "B" safeguards room sumps have 2 sump pumps per flow rate of 50 gpm at 10C ft. of each with a design operating The The sump ptaps are powered via non-vital MCCs 1A2/1B2.
safeguards high and high-high sump alarms are powered from vital head.
l i
MCC 1AB.
l The function of the safeguards pump room surp isolation valves from RAB is to isolate the safeguards rooma (HCV-25-1 through 7)to a fire main break (FSAR Section 9.5A-110).
The i
safeguards pump room sump isolation valves are normally open' to flooding due l
allow drainage to the ECCS sump.
j 4
Analysis of Eff ects of Lif tints V3439 i
4.0 The lifting of V3439 will occur while on SDC if LPSI pump operation relief discharge header pressures reaching theSufficient pressure to results in LPSI valve setpoint (500 psig i 34).
relief valve may occur during LPSI pump starts that result in aThe pr- -".re spike (m30 psi).sps AP of 182 poi ud a suction pressure of betwe 268 psia, will reach the minimum design setpoint.
With the ti..
blowdown for the valve set at 14%, the valve will continue to relieve until system pressure drops below approwNataly 430 peig.
Therefore, the analysis operation thst involves LPSI discharge ~ pressures approach relief valve setpoint. relief valve setpoint during operation with the shutdow system isolated from the RCS.)
Loss of RCS Inventory If V3439 lifts during normel shutdown cooling (SDC) o valve ressats.
Since the flow rate through the valve at the setpoint pressure is limited to about 40 gpa, the rate of inventory loss is well within normal charging system capability.
During long term cooling following a design hamia accident the SDC system may be used to satisfy the RCS heat removal safety function Assuming the SDC system is being utilised, then the of 40 pressure safety injections pumps or charging pumps.
4
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JPN-PSL-SENP-95-101 REVISION 0 l
PAGE 6 OF 8
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Loss of RFP Inventerv j.
During the injection phase of small break loss of coolant acciderts (IDCAs), excess steam demand events (ESDE.s) and steam generator tube rupture (SGTRs) with LPSI pumps aligned to the refueling water
+
l tank (RWT), LPSI pump maximum discharge pressure is about 230 psig j
(RWP head plus LPSI pump ap). This is well below the lif t setpoint of V3439. Therefore a loss of RFP inventory vould r.ot'c: w.
1 t
1 Loss of Containment Sumn Inventor's During the recirculatior phase of LOCAs, the LPSI disdarge header This is pressure will no.t approach the lift se'.: point of V3439.
because the LPSI pumps are either autc%atically stopped on receipt
)
of recirculation actuation signal, or if used for hot leg injection, maximum LPSI discharge header pressure would be 2 260 i
l psia (containsant sump head plus LPSI pump ap), well below the lif t j
setpoint of V3439. Therefore a loss of containment sump inventory would not occur.
Floodina in the Pine Tunnel l
During the subject event the pipe tunnel floor drains were isolated f
which caused flooding in the pipe tunnel up to the height of the lip on the door (mlo"). Reactor coolant then spilled into the RAB i
hallway where it was than routed to the "B" safeguards room sump via an RAB floor drain. Se lower levels of the pipe tunnel do not
'in electrical equipment that is susceptible to faults from a tacs. ding event of'this type.
When the safeguards pump room sump isolation valves att open, the discharge of V3439 is directed to the floor drains la the RAB pipe tunnel at the
.5 ft. elevation.
These drains empty into the "B" safeguards room sump via normally opened air operated valves controlled from a k. Ole switch on RNB 106. D e capacity of these drains exceeds the 40 gym relief discharge, and therefore, the pipe tunnel will not flood if the drains are available.
Floodina in the "B" Safeauards Pn=n Room The "B" safeguards room nunp holds 1100 gallons and automatically pumps down to the equipment drain tank (EDT) at = 850 gallons at a rate of 50 gpa (capacity of one sump pump).
If level reaches 950 gallons, the second sump pump starts. Assuming non-vital power to
-the pumps is maintained, flooding of the "B" safeguards room will Assuming a loss of offsite power (i.e., no safeguards not occur.
sump pumps available), and that the di=>*=W water is undetected (i.e., no operator intervention), the water level in the safeguards would reach the "B"
EPSI pump motor in m 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
room Annunciation of the safeguards high and high-high sump alarms would require investigation to determine the reason for the alarm (Ref 3).
5"ven hours is considered sufficient time to identify and isolate the leak prior to loss of the "B" EPSI pump.
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M JPN-PSL-SENP-95-101 f fgI s
REVISION O L
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Radiolocical Consecuences of Damian Basis Accidents s
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Increases e s ESDEs and SGTRs are the design basis accidents v
4 mall break LOCAs, that could involve lif ting of V3439 since these events could result
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M in the SDc system being placed in service with maximum suction Q % pressure (i.e., 268 psia) supplying the LPSI pump (s).
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SAR analysis of small break LOCAs, ESDEs and SGTRs demonstre.te these events involve limited fuel damage and M'nlogical sequences when compared to the large break LOCA an n ysis (i.e.,
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for whole body & thyroid dose).
the limiting event
' Assuming thac the flocr drains are not isolated durf ig one of these l
E accidents, air-borne radioactive materials will be filtered out by For this case, which l
N<?" the safety rela %d ECCS ventilation system. consequences are estimated to considered to be the most likely, I
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'E ease by a very small fraction and will remain well below the p kh 4 limiting large break LOCA analysis.
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l, p[h f normal pipe tunnel drainage is not available and leakage g e
hallway, some additional air-borne radioactive materials will beThe drai hg O
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\\.<9 g
[jsd 0 exhausted to the atmosphere. surveillance on 7/10/95 (32 days).
r4 p\\ /M(MBased on the baseline Probabilistic Safety As======
ksince the last valve stroke (PSA), the l
- total core damage frequency (CDF) contribution from the equivalent w
J (small-small and small LOCAs, SCTRs, and k' ' 4 $ PSA accident scenarios l
Main Steamline Breaks) is 1.0E-5/yr. 'the core damage probability j
- J 3P) over the 32 days that the drains were assumed unavailable is 1'$\\,.
The s.sE-7 represents the probability.over, 4
.stimated to de s.sa-7.
^ ir-born [e y 9 \\a @e_32_. day _ period _oLexhausting _a_small_30antity" t
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g CDP "lis,clo 1
ed insign t.
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Tf R&' &ct_ bewn WY N'
coNCtUSI NE =~S R k Isr n = 'dW bM *b
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The safety significance of Alf ing V3439 with repect to plant j
e effects of loss of inventory, operation has been evaluatRAB flooding and radiological c nsequenc i
j the lifting of V3439 would not have a significant effect on safe plant operation during normal, shutdown and design concluded that f
basis accident conditions.
l v==IFIc1 TION 8DMM1MY J
6.0 The scope of this verification was to review the inputs 24 The method used for thit determine if the results were reasonable.
verification consisted of ensuring that the applicable references, codes, and regulatory require-ents were identified and addressed i
l The inputs are correctly selected and applied.
J The conclusions provided are reasonable with respect to the input and discussions.
The verifier concurs with the Nuclear Safet
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i JPu-PsL-SENP-95-101 j
REvISzou o PAGE 8 OF S Related classification of this Engineering Evaluation.
The was verified l
rationale in assigning the safety classification l
against the requirements of JPN Quality Instructions. The verifier j
i j
concurs with the conclusions outlined above.
l i
b 7.O REF2RE3fCES In-House Event IHE-95-046, dated 8/17/95.
St Lucie Unit 1 Emergency Operating Procedure 1-EOP-03, Loss l
1) 2)
of Coolar.t Accident, Rev 12.
1-St Lucie Unit 1 Ilant Annunciator Summary Produre ONP l
l 3) 0030131, Rev 60 (Nindow R-14) l l
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St Lucie Unis 1 FSAR, amend. 14.
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