REGULATOR INFOR»'»iAT'ION O'ISTRI BUT IONi TEN (RIDS)

DOCKEiT< 0:05000335

NOTES:

ACCESSION NBR: 820908021,8'OC>. DATE'2/08'/31, NOTARIZED: NO

FACIL-'50-335 St. L4cie Pl a'nt'i .Unit ii Ftlor ida'ower 8 Liight Cb».

AUTH.NAME» AUTHOR AFFCLIIAT»IONUHRIGi R E" Florida Power. E, L,ight» Co..

RECIP ~ NA»ME< RECiIPiIEi»l»Tl AFFILIIA)ION»,EISENHUT<O.G. Dii vi si on» os~ Lli censino

SUB JECiT: For wards„a'ddli anall uses re reactor cool ant~ oumo«„sei zed rotorw/loss of» of fai te, oozier w/worst single active failure.8, lossof ail'1r nonemeroencv- a'c. power w/wor st single active- fai lure<oer- 81,1118 Lit r L~8'1."480 ~

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P. OX 14000, JUNO BEACH, FL33408

FLORIDA POWER & LIGHT COMPANY

August 31, 1982L-82-381

Office of Nuclear Reactor RegulationAttention: Mr. Dar rell G. Eisenhut, Director

Division of LicensingU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

Dear Mr. Eisenhut:

Re: St. Luci e Unit 1

Docket No. 50-335Stretch Power - Additional Analysis

Please find attached the following analyses for St. Lucie Unit. 1:

(a) Reactor coolant pump seized rotor with loss of offsitepower with the worst single active failure (attachment1);

(b) Loss of'll non-emergency AC power with the worstsingle active failure (attachment 2).

These analyses are submitted as committed by our .letter L-81-484, datedNovember 18, 1981, and represent the final outstanding information request ofour Stretch Power Application.

Our re-evaluation of these events demonstrated acceptable results. .Theseanalyses were conducted using conservative assumptions and the single activefailure which in our engineering judgment was the most limiting event.

Very truly yours,

Robert E. UhrigVice PresidentAdvanced Systems 5 Technology

REO/JEM/mbd

Attachment

cc: Mr. James P. O'Reilly, Region IIHarold F. Reis, Esquir e

sa09Oa02>a 820SSSPDR ADOCK 05000335P PDR

IIOl

PEOPLE...'SERVING PEOPLE

ATI'ACHMENT 1

Introduction

The Seized Rotor event was reanalyzed for Cycle 5 to include loss of offsitepower following turbine trip, and one atmospheric dump valve (ADV) stuck openat the t ime of the event initiation. 'Ibis additional single failure wasconservatively included in the analysis even though 'the ADVs are not allowedto operate in the automatic mode. Since the ADVs'etopen pressure is belowthe setpoint for the main steam safety valves (MSSVs), the ADVs may bechallenged during the secondary pressurization produced by the loss of thecondenser on loss of offsite power. One of the two ADVs is assumed'o stickopen due to mechanical failure.

'Ihe single reactor coolant pmp shaft seizure is postulated to occur as aconsequence of a mechanical failure. Following this, the reactor coolant flowstarts to decrease. A reactor tr ip is initiated by a low coolant flow rate, asdetermined by a reduction in the sun of the individual. loo~ coolant

signals. This occurs when the flow rate decreases to 93 percent ofthe initial flow.

Discussion

The initial conditions for the Seized Rotor event are listed in Tables 1 and 2,and are consistent with the initial conditions assumed in Reference 1. Otherassmptions on key parameters are listed below:

A. The NSSS response is simulated by CESEC, a digital computer code describedin Reference 2.

B. Upon initiation of this transient, core flow is .modeled to start, tocoastdown due to the resistance to forced flow by the seized pump. The

:. flow decreases rapidly at first, and then approaches the asymptotic three'ump core flow value about 1.5 seconds'fter initiation of the transient.

C. One of the two ADVs is conser vatively assumed to stick open due tomechanical failure at time zero. The other ADV is assumed to be onautomatic mode.

D. The analysis assumes a 3.0 second delay time between the time of turbinetrip and the time of loss of offsite power.

E. The auxiliary feedwater flow is assumed to initiate automatically, with adelay of 120.0 seconds after the steam generator level setpoint of 25.5 ft,above the tube sheet is reached. 'Ibis delay accounts for system response",

, time.

F. The RCS flow coastdown, the limiting axial power distribution for the mostnegative axial shape index allowed within the full power shape index LCO,and a consistent scram reactivity curve are input to STRIKIN II (seedescription in Reference 3) to 'determine the hot channel'nd core averageheat fluxes versus time during the tr ansient.

0TORC/CE-1 is used to calculate the minimum DNBR .fer the transient; TheSeized Rotor transient is initiated at the Limiting Conditions forOperation to determine the minimum DNBR.

H. ln determining the predicted number of fuel pin failures, the TORC code ised t al lat the DNBR versus radial peaking factor. An integral fuel

damage'alculation is then carried out by combinxng the resul ts from TORC'ith the number of fuel rods having a given radial peaking factor. Thenumber of fuel rods versus radial peaking factor is. taken from a cumulativedistribution of the fraction of fuel rods with nuclear radial peakingfactor in a given range. 'Ibis yields a distribution Gf the fraction ofpins with a particular DHBR as a function ef DNBR. This information isthen convoluted with a probability of burnout vs. DNBR to obtain the amountof. fuel failure. This method is discussed in detail in CEHPD-183, "C-EHethods for Less ef Flow Analysis" (Reference 5). lt is totally consistent,with the methed described in that topical report and with methodspreviously used and approved for St. Lucie Unit 1, Cycle 5 (Reference 1).

X. A conservatively "flat" pin census distribution (a histogram of the numberof pins with radial peaks in intervals of 0.01 in radial peak normalized tothe maximum peak) is used to determine the number of pins that experienceDNB.

J. The,major portion of the radiological releases resulting from this eventcombination is from the activity released by the failed fuel into theprimary coolant. All of the activity which enters the secondary coolant,,through steam gener ator tube leakage, is conser vatively assumed to be

. re3,eased to the atmosphere (ne credit is taken for iodine partitioning inthe steam gener ators). See Table 3 for the assumptions used forcalculating the radiological release te the atmosphere.

K, Qe operators are assumed to begin controlling 'the plant cooldown at 30minutes, using the operable ADV and closing the ADV block valve to isolate

* .the stuck open ADV.

Table 4 gives a chronological list of system actions and relevant plantparameter values for the Seized Rotor event, initiated from an axial shape indexval0e'f -0. 11. Figures 1 through 5 show core power, core average heat flux,RCS coolant temperatures, RCS pressure and S.G; pressure during the tr ansient.The minimum DNBR occurs at 1.7 seconds after transient initiation. This timeis well before the time of less of offsite power at 3.82 seconds. Therefore,the minimum DHBR for this event is roughly equivalent as that for a lockedrotor without loss of offsite power. The number ef,fuel pins predicted to failis equal to 1.63$ . 'Ihe RCS pt essure reaches a maximum value of 2412.40 psia at,7.4 seconds.

fTable 5 lists the steam releases via the ADVs and the main steam safety valves{MSSVs) which were calculated for this transient. Based on'he r eleases, the .J y

0-2 Hrs site boundary doses are:

Thyroid (DEQ I-131): 36.1 REM

Mhole Body (DEQ Xe-133): 0.06 REM

Conclusion

The evaluation shows that the plant response to a one pump r esistance to forcedflow (shaft seizure) with a loss of offsite power, Technical Specificationstean generator tube leakage and one stuck ep n a~spheric dump valve resultsin a small fraction of fuel pins experiencing failure. The corre'sponding siteboundar y dose is within the 10CFR100 acceptance guidelines. In addition, themaximum RCS pressur e experienced during the event is well under . the upset,pressure limit of 2750 psia.

I

~ ~

TABLE 1

SEIZED ROTOR WITH LOSS OF AC AND STUCK OPEN ADV .

KEY PARAMETERS ASSUMED IN PIN FAILURE CALCULATIONS

Paraneter

Initial Core Power Level

Initial Coolant InletTemperature

Initial Core Mass Flow Rate

Reactor Coolant System Pressure

Moderator Temperature Coefficient

Doppler Coeffici:ent Multiplier

CEA North on Trip

Integrated Ra(ial Peaking Fasterwith Tilt, Fr

P

Axial Shape Index

Low Flow Analysis Trip

Units

t%t

OF

10 ibm/hr

psia

x10 ~f F

5 of initial flow

2700

549

138. 3

2225

+0.5

0.85

-5e6

1. 70

~ ~

'0.

11

93. 0

Cycle 5

e

TABLE 2

SEIZED ROTOR WITH LOSS OF AC AlG) STUCK OPEN ADV .

KEY PARAMETERS ASSUMED IN STEAM RELEASE CALCULATIONS

-Paraneter

Initial Core Power Level

Initial Coolant InletTenperature

Initial Core Mass Flow Rate

Initial Reactor Coolant SystemPressure

Initial Steam Generator Pressure

Initial Steam Generator Level

Low Flow Analysis Trip Setpoint

Moderator Temperature Coefficient

Doppler Coefficient Feltiplier

CEA North on Trip

Reactor Regulating System

Steam Bypass System

Auxiliary Feedwater System

Units

OF

10 lhn/hr

psla

psia

5 of initial flow

x10 ~ MF

Cycle 5

551

138. 3

2200

f00

36.2 abovetube sheet

93 0

0. 85

-5.6

Manual Mode

Inoperative

Automatic

~ e ~ de

TABLE 3

SEIZED ROTOR lGTH LOSS OF AC AND STUCK OPEN ADVKEY PARAMETERS ASSUMED IN THE RADIOLOGICAL EVALUATION

Par arne te)

Primary to Secondary Leak Rate 1

Reactor Coolant System Volume{Excluding Pressurizer)

Reactor Coolant System Maximum'Allo~able Concentration (DEQ I-131)"

Steam Generator Haximum glouable. Concentration (DEQ I-131)

Reactor Coolant'System MaximumAllowable Concentr~tion of NobleGases (DEQ Xe-,133)

Steam Generator Partition Factor

Atmospheric Dispersion Coefficient 2

Breathing Rate

Dose Conversion Factor (I-131)

Gaits

GPM

p Ci/gm

p Ci/gm

uCi/gm

sec/M3

M3/sec

REM/Ci

Value

1.0

'9601

1.0

0.1

100

E-

1.0

8.55x10 .5

3. 47x10

1. 48x10

1 Tech Spec limits2 0 - 2 hour accident condition

d~ d

e

\

TABLE 4~ g

SEIZED ROTOR MITH LOSS OF AC AHD STUCK OPEN ADVSEQUENCE OF EVEHTS

Time (sec)

0.0

0.0

0. 17

0.82

0.82

1-32

1-33

2 13

3. 26

3.82

4.50

7.4

».6

12.6

39-9

40. 8

1056.7

1390.0

1420.0

Event

Seizure of RC Pump Shaft

Inadvertent Opening ef ADVin Affected Loop

Low Flow Septoint Reached

Reactor Trip Signal Generated

Turbine Trips

Rods Dr opped

Maximum Power

Opening of ADV in UnaffectedLoop

Unaffected Loop Main SteamSafety Valves Open

Loss of.Offsite Power

Main Steam Safety Valves Open,Affected Loop

Maximum RCS Pressure

Unaffected LeepMaximum S.G. Pressure

Affected LoopMaximum S.G. Pressure

ADV in Unaffected Loop Closes

MSSVs in Affected Loop Close

Auxiliary Feedwater Begins toEnter Affected S.G.

SIAS is Actuat.ed

Safety Inje'ction Pumps ReachFull Speed

Set int or Value

93$ of init,'ial4-pump flow

105.14'70

psia

1000 psia

1000 psia

2412 40 psia

1078.74 psia

1055.23 psia

970 psia

961.2 psia::;;..e

125.4 lb/second 'I

1578 psia

\ ~ ~

TABLE 4(continued)

1~33 9

1503 7

. 1800.0

8561.0

Pressurizer Empties

Safety Injection Flm Starts

Operator Takes Control of AvailableADVs to Initiate Plant CooldownOperator Closes Affected ADVBlock Valve

Shutdown Cooling Initiated;RCS Average Temperature

325 F

~II

TABLE 5

SEIZED ROTOR WITH LOSS OF AC AND STUCK OPEN ADV'TEAM RELEASES

Xnte ated Steam Release

Steam Release Through Safety Valves'During 0 — 2 hrs, ibm

Steam Release Through Ataaspheric Steam ImpValves During 0 - 2 hrs, ibm

Total Amount of Steam Released During0 - 2 hrs, ibm

Total Amount of Steam Released Until Shutdown Cooling's

Initiated (325 F), ibm

Va1ue

1490]

750448

765349

903094

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FLORD I A'ONER 4t LIGHT

CO't.

Lucre<uclear Power Plant

~ SEIZED ROTOR KITH LOSS OF AC AND STUCK OPEH ADV

CORE I'Ol<ER YS TINE

FIGURE

j.

120

100

I—

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60

U

LU

40

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20

0

0

TIiNE, SECOf<DS

16 -20

FLOR I DAHER 5 LIGHT COs

St. LucreNuclear Power Plant

SEIZED ROTOR MITH LOSS OF AC AND STUCK OPEN ADV

COPE HEAT FLUX YS TIWE

FIGURE

700

600o

500

400

5c '00

200

TAVG

1000 400 . Soo 1200

TINE, SECONDS

1600 1800

FLORIDAPOWER 5 LIGHT CO>

St. LucreNuclear Power Plant

SEIZED ROTOR WITH LOSS OF AC AND STUCK OPEN M3V

REACTOR COOLANT SYSTEN TENPERATURES VS 7IHEFIGURE

3 *

2400'000

1600

1200

400

0 400 800. 1200- TINE, SECONDS

1600 1800

FLOR I DANER E( LIGHT COe

St. Lucrenuclear Power Plant

SEIZED ROTOR HITH LOSS OF AC AND STUCK OPEN RVREACTOR COOLANT SYSTEN PRESSURE VS TINE

FIGURE

1200

1000

800

600

400

200

0

400 8QO j200

TINE, SECONDS

1600 1800

FLORIDAHER h LIGHT COi

St. I ucieuclear Power Plant

SEIZED ROTOR METH LOSS OF AC AND STUCK OPEN ADV.

~ STEAN GENERATOR PRESSURE YS TINEFIGURE

5

ATTACHMENT 2

~troduction

The Loss of, Offsite Power event was reanalyzed for Cycle 5 to include one.atmospheric dump valve (ADV) stuck open at the time of the event initiation.This additional single failure was conservatively included in the analysis eventhough the AD& are not allowed to operate in the automatic mode. Since ADV

setopen pressure is below the setpoint for the main steam safety valves(MSSVs), the ADVs may be challenged during the secondary pressurizationproduced by ttfe loss of the condenser on loss of offsite power. One of theADVs'is assumed to stick open due to .mechanical failure.

We event is initiated . by a loss of all non-emergency AC power. Followingthis,. the reactor coolant flow starts to decrease. A reactor trip is initiatedb a low ooolant flow rate, as determined by a reduction in the sun of they a o osteam generator differential pressure signals. This occurs when the fldecreases to 93 percent of the initial flow.

Discussion

The initial conditions for the event are listed in Table 1 and are consistentwith the initial conditions assumed in Reference 1. Other asstxnptions on keyparameters are listed below:

A. We NSSS response is simulated by CESEC, a digital computer code describedin Reference 2.

B. Upon initiation of this transient,, core flow is modeled to start tocoastdown due to the loss of AC power.

d r g

Table 0 lists the steam releases via the ADVs and the main steam safety valves(MSSVs), which were calculated for this transient. Based on the releases, the0-2 Hrs site boundary doses are:

C. Cne of the ADVs is conservatively assumed to stick open due to mechanical,'failure at time zero. The other ADV is assumed to be in automatic mode;'.'e operators are assed to begin controlling the plant cooldown at 30minutes, using the operable ADV and the auxiliary feedwater, and closingthe ADV block valve to isolate the stuck open ADV.

For the first few seconds of the transient, the Loss of Offsite Power event,behaves l'ike a complete Loss of. Forced Reactor Coolant Flow event. The minimumDNBR limit was not exceeded and no fuel failure was predicted for that event.Thus, the DNBR limit will not be exceeded for the Loss of Offsite Power event.

Table 3 gives a chronological list of system actions and relevant plantpar'ameter values for the event. Figures 1 through 5 show core power, coreaverage heat flux, RCS coolant temperatures, RCS pressure and S.G. pressur'e

u in the transient.

Wyroid (DEQ I-131): 1.53 REM

Mhole Body (DEQ Xe-133): 0.0009 REM

0 ~The results of the analysis shows that the peak RCS. pressure is 2457 psia whichoccurs at 3.8 seconds.

Conclusion

The evaluation shows that the plant response to a loss of,offsite power',Technical Specification steam generator tube leakage and one'tuck openatmospheric dump valve results in a maximum offsite dose which is within the10CFR100 acceptance guidelines.,

I ~ e

~ ~ ~I

~ ~

4 eTABLE 1

LOSS OF AC AND STUCK OPEN ADVKEY PARAMETERS ASSUMED IH STEAM RELEASE CALCULATXOHS

Parameter

Initial Core Power Level

Units ~Cele 5

Xnitial Coolant InletTemperature

Initial Core Mass Flow Rate

Xnitial Reactor Coolant SystemPressur e

Initial Steam. Generator Pressure

Initial Steam Generator Level

Low Flow Analysis Trip Setpoint

Moderator Temperature Coefficient

Doppler Coefficient Multiplier

CEA North on Trip

Reactor Regulating System

Steam Bypass Systen

'uxiliary Feedwater System

OF

10 lhn/hr

psia

psia

5 of initial flow

x10 bp/ F

551

138. 3

2300

900

36.2 abovetube sheet

93 0

«0,5

0.85

-5.6

Manual Mode

Inoperat ve

Automatic

s

Parameter

. ~TABLE 2

LOSS. OF AC AND STUCK OPEN ADVKEX PARAMETERS ASSUMED IN THE RADl'OLOGl'CAL EVALUATION

Units Value

Primary to Secondar.y Leak Rate

Reactor Coolant System Volume(Excluding Pressurizer)

Reactor Coolant System MaximumAllovab3,e Concentration (DEQ I-131)

Steam Generator Maximum plowableConcentration (DEQ I-131)

Reactor Coolant System MaximumAllowable Concentr~tion of NobleGases (DEQ Xe-133)

Stean Gener ator Par t ition Factor

Atmospheric Dispersion Coefficient

Breathing Rate

Dose Conversion Factor (I-131)

Ft3

u Ci/gm

g Ci/gm

.p Ci/gm

sec/M3

M3/sec

REYi/Ci .

1.0

9601

1.0

0.1

100

1.0

8.55x10 5

3.<7x10-"

1. 48x106

Tech Spec limits

0 — 2 hour accident condition

TABLE 3

~ Time (sec)

0.0

0.0

0.86

0.88'-88

2.03

2. 08

2.5

3.28

3. 29

3.80

13.38

41.2

1186..1

1592.6

1622.6

LOSS. OF AC AND STUCK OPEN ADVSEQUENCE OF EVENTS

Event

Loss of Offsite Power

Inadvertent Opening of ADVin Affected Loop

Low Flow Setpoint Reached

Reactor Trip Signal Generated

Turbine Trips

Maximu'm Power

Opening of ADV in UnaffectedLoop

Rods . Dr opped

Unaffected LoopMain Steam Safety Valves Open

Affected LoopMain Steam Safety Valves Open.

Maximum RCS Pressure

Maximum S.G. Pressure

Unaffected Loop, ADV Closed

Auxiliary Feedwater Begins toEnter Affected S.G.

Safety Injection ActuationSignal Generated

Safety Injection Punps ReachFull Speed

~ Set, int or Value

93+ of in'i,tial4-pump flow

103.7$

970 psia

1000 psia

1000 psia

2457.0 psia

1093.5 psia

970 psia

125.4 ibm/sec

1578 psia

1800.0

8640.0

Operator Takes Control of AvailableADV to Initiate Plant, CooldownOperator Closes Affected ADVBlock Valve

Shutdown Cooling Initiated;RCS Average Temperature

~~

~ ~

TABLE 4

Inte ated Steam Release

LCSS OF AC AND STUCK OPEN ADVSTEAM RELEASES

Value

Steam Release. Through Safety ValvesDuri'ng 0 - 2 h.s, lhn

Steam Release Through Atmospheric SteamEh'alvesDuring 0 - 2 hrs, ibm 4

Total Amount of Steam Released During0'- 2 hrs, ibm

Total Amount of Steam Released Until Shutdown Coolingis Initiated (325 F), ibm

]634]

742712

759053

913780

~ a

* ~ ~

P

~ ~

l—'0CDCD

LL.CD

QQ

CDCL

40

. 0 400 800 1200 1600 1800

TINE, SECONDS

F LOR I DAPOHER S LIGHT CO

St.. LvcieNuclear Power Plant

'LOSS OF OFFSITE POHER EVENT

CORE POHER VS TINEF IGURE

~ + ~

'100

C3CO

.80

20

400 800 1200

TINE, SECONDS

it-00 ZSOO .

aI.

FLORIDAOHER g LIGHT CO

St. Lucieuclear Power Plant

LOSS OF OFFSITE PONER. EYENT

CORE HEAT FLUX YS TINE

F I GURE

2

. 700

600O

TOUT

TAVG

500TIN

400

~ CD.

'.'

200

800 1200

TINE, SECONDS

1S00 1800

FLORIDA'ONER & LIGHT CO<

St. LUcienuclear Power Plant

LOSS OF OFFSITE POHER EVENT

REACTOR COOLANT SYSTEi'1 TEMPERAT'ORES.VS T'INEFIGURE

2400

2000

.}600

}200

800

400

400 800 }200}

} 6»j.SOO

TINE, SECONDS

FLORIDAONER 5 I ICjHT COi

St. Lucre'uc1ear Power. P'lant

LOSS OF OFFSITE POWER EVENT

REACTOR COOLANT SYSTEfl PRESSURE VS TINE~

~

FIGUREI}

0 ~ 'T.

1200

.800

600

400

200

00 400 800 . 1200

TINE, SECONDS

1600 1800

FLORIDAWER 5 LI,GHT COi

St. Luci ecl ear Power Plant

LOSS OF OFFSITE POWER EYENT

STEArl GENERATOR PRESSURE YS TIj1E

FIGURE

5

'0 Nl

V

l

I