L-82-381, Forwards Addl Analyses Re Reactor Coolant Pump Seized Rotor W/Loss of Offsite Power W/Worst Single Active Failure & Loss of All Nonemergency Ac Power W/Worst Single Active Failure, Per L-81-484

Forwards Addl Analyses Re Reactor Coolant Pump Seized Rotor W/Loss of Offsite Power W/Worst Single Active Failure & Loss of All Nonemergency Ac Power W/Worst Single Active Failure, Per L-81-484

ML17213A404
Person / Time
Site:	Saint Lucie
Issue date:	08/31/1982
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-82-381, NUDOCS 8209080218
Download: ML17213A404 (34)

Text

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>>ION UHRIGi R E"

Florida Power.

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EISENHUT<O.G.

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For wardsa'ddli anall uses re reactor cool ant~ oumo<<sei zed rotor w/loss of>> of fai te, oozier w/worst single active failure.8, loss of ail'1r nonemeroencv-a'c. power w/wor st single active-fai lure<

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OX 14000, JUNO BEACH, FL33408 FLORIDAPOWER & LIGHT COMPANY August 31, 1982 L-82-381 Office of Nuclear Reactor Regulation Attention:

Mr.

Dar rell G. Eisenhut, Director Division of Licensing U.S.

Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Eisenhut:

Re:

St.

Luci e Unit 1

Docket No. 50-335 Stretch Power - Additional Analysis Please find attached the following analyses for St. Lucie Unit. 1:

(a)

Reactor coolant pump seized rotor with loss of offsite power with the worst single active failure (attachment 1);

(b)

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

These analyses are submitted as committed by our.letter L-81-484, dated November 18,

1981, and represent the final outstanding information request of our Stretch Power Application.

Our re-evaluation of these events demonstrated acceptable results.

.These analyses were conducted using conservative assumptions and the single active failure which in our engineering judgment was the most limiting event.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 5 Technology REO/JEM/mbd Attachment cc:

Mr. James P. O'Reilly, Region II Harold F. Reis, Esquir e sa09Oa02>a 820SSS PDR ADOCK 05000335 P

PDR IIOl PEOPLE...'SERVING PEOPLE

ATI'ACHMENT 1 Introduction The Seized Rotor event was reanalyzed for Cycle 5 to include loss of offsite power following turbine trip, and one atmospheric dump valve (ADV) stuck open at the t ime of the event initiation.

'Ibis additional single failure was conservatively included in the analysis even though 'the ADVs are not allowed to operate in the automatic mode.

Since the ADVs'etopen pressure is below the setpoint for the main steam safety valves (MSSVs),

the ADVs may be challenged during the secondary pressurization produced by the loss of the condenser on loss of offsite power.

One of the two ADVs is assumed'o stick open due to mechanical failure.

'Ihe single reactor coolant pmp shaft seizure is postulated to occur as a

consequence of a mechanical failure.

Following this, the reactor coolant flow starts to decrease.

A reactor tr ip is initiated by a low coolant flow rate, as determined by a reduction in the sun of the individual. loo~

coolant signals.

This occurs when the flow rate decreases to 93 percent of the 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.

Other assmptions on key parameters are listed below:

A.

The NSSS response is simulated by

CESEC, a digital computer code described in Reference 2.

B.

Upon initiation of this transient, core flow is.modeled to

start, to coastdown 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 to mechanical failure at time zero.

The other ADV is assumed to be on automatic mode.

D.

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

E.

The auxiliary feedwater flow is assumed to initiate automatically, with a delay 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 most negative axial shape index allowed within the full power shape index

LCO, and a

consistent scram reactivity curve are input to STRIKIN II (see description in Reference

3) to 'determine the hot channel'nd core average heat fluxes versus time during the tr ansient.

0 TORC/CE-1 is used to calculate the minimum DNBR.fer the transient; The Seized Rotor transient is initiated at the Limiting Conditions for Operation to determine the minimum DNBR.

H.

ln determining the predicted number of fuel pin failures, the TORC code is ed 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.

The number of fuel rods versus radial peaking factor is. taken from a cumulative distribution of the fraction of fuel rods with nuclear radial peaking factor in a given range.

'Ibis yields a distribution Gf the fraction of pins with a particular DHBR as a function ef DNBR.

This information is then convoluted with a probability of burnout vs.

DNBR to obtain the amount of. fuel failure.

This method is discussed in detail in CEHPD-183, "C-E Hethods for Less ef Flow Analysis" (Reference 5). lt is totally consistent, with the methed described in that topical report and with methods previously used and approved for St. Lucie Unit 1, Cycle 5 (Reference 1).

X.

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

J.

The,major portion of the radiological releases resulting from this event combination is from the activity released by the failed fuel into the primary 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 in the steam gener ators).

See Table 3

for the assumptions used for calculating the radiological release te the atmosphere.

K, Qe operators are assumed to begin controlling 'the plant cooldown at 30

minutes, 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 plant parameter values for the Seized Rotor event, initiated from an axial shape index val0e'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 time is 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 locked rotor without loss of offsite power.

The number ef,fuel pins predicted to fail is equal to 1.63$.

'Ihe RCS pt essure reaches a maximum value of 2412.40 psia at, 7.4 seconds.

f Table 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 forced flow (shaft seizure) with a loss of offsite

power, Technical Specification stean generator tube leakage and one stuck ep n a~spheric dump valve results in a small fraction of fuel pins experiencing failure.

The corre'sponding site boundar y dose is within the 10CFR100 acceptance guidelines. In addition, the maximum RCS pressur e experienced during the event is well under

. the

upset, pressure limit of 2750 psia.

I

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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 Inlet Temperature Initial Core Mass Flow Rate Reactor Coolant System Pressure Moderator Temperature Coefficient Doppler Coeffici:ent Multiplier CEA North on Trip Integrated Ra(ial Peaking Faster with 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 Inlet Tenperature Initial Core Mass Flow Rate Initial Reactor Coolant System Pressure 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 above tube sheet 93 0

0. 85

-5.6 Manual Mode Inoperative Automatic

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~ de TABLE 3 SEIZED ROTOR lGTH LOSS OF AC AND STUCK OPEN ADV KEY 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 Maximum Allowable Concentr~tion of Noble Gases (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 limits 2 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> accident condition d

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TABLE 4

~ g SEIZED ROTOR MITH LOSS OF AC AHD STUCK OPEN ADV SEQUENCE 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 ADV in Affected Loop Low Flow Septoint Reached Reactor Trip Signal Generated Turbine Trips Rods Dr opped Maximum Power Opening of ADV in Unaffected Loop Unaffected Loop Main Steam Safety Valves Open Loss of.Offsite Power Main Steam Safety Valves

Open, Affected Loop Maximum RCS Pressure Unaffected Leep Maximum S.G. Pressure Affected Loop Maximum S.G. Pressure ADV in Unaffected Loop Closes MSSVs in Affected Loop Close Auxiliary Feedwater Begins to Enter Affected S.G.

SIAS is Actuat.ed Safety Inje'ction Pumps Reach Full Speed Set int or Value 93$ of init,'ial 4-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

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TABLE 4 (continued) 1~33 9

1503 7

. 1800.0 8561.0 Pressurizer Empties Safety Injection Flm Starts Operator Takes Control of Available ADVs to Initiate Plant Cooldown Operator Closes Affected ADV Block 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 Imp Valves During 0 - 2 hrs, ibm Total Amount of Steam Released During 0 - 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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I ucie uclear Power Plant SEIZED ROTOR METH LOSS OF AC AND STUCK OPEN ADV.

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STEAN GENERATOR PRESSURE YS TINE FIGURE 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 even though 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 pressurization produced by ttfe loss of the condenser on loss of offsite power.

One of the ADVs'is assumed to stick open due to.mechanical failure.

We event is initiated. by a loss of all non-emergency AC power.

Following this,. the reactor coolant flow starts to decrease.

A reactor trip is initiated b

a low ooolant flow rate, as determined by a reduction in the sun of the y

a o

o steam generator differential pressure signals.

This occurs when the fl decreases to 93 percent of the initial flow.

Discussion The initial conditions for the event are listed in Table 1

and are consistent with the initial conditions assumed in Reference 1.

Other asstxnptions on key parameters are listed below:

A.

We NSSS response is simulated by

CESEC, a digital computer code described in Reference 2.

B.

Upon initiation of this transient,,

core flow is modeled to start to coastdown 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, the 0-2 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 30

minutes, using the operable ADV and the auxiliary feedwater, and closing the 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 minimum DNBR 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 plant par'ameter values for the event.

Figures 1 through 5

show core

power, core average 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 which occurs 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 open atmospheric dump valve results in a

maximum offsite dose which is within the 10CFR100 acceptance guidelines.,

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TABLE 1 LOSS OF AC AND STUCK OPEN ADV KEY PARAMETERS ASSUMED IH STEAM RELEASE CALCULATXOHS Parameter Initial Core Power Level Units

~Cele 5

Xnitial Coolant Inlet Temperature Initial Core Mass Flow Rate Xnitial Reactor Coolant System Pressur 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 above tube sheet 93 0

<<0,5 0.85

-5.6 Manual Mode Inoperat ve Automatic s

Parameter

~

TABLE 2 LOSS. OF AC AND STUCK OPEN ADV KEX 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 Maximum Allovab3,e Concentration (DEQ I-131)

Steam Generator Maximum plowable Concentration (DEQ I-131)

Reactor Coolant System Maximum Allowable Concentr~tion of Noble Gases (DEQ Xe-133)

Stean Gener ator Par tition 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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 ADV SEQUENCE OF EVENTS Event Loss of Offsite Power Inadvertent Opening of ADV in Affected Loop Low Flow Setpoint Reached Reactor Trip Signal Generated Turbine Trips Maximu'm Power Opening of ADV in Unaffected Loop Rods. Dr opped Unaffected Loop Main Steam Safety Valves Open Affected Loop Main Steam Safety Valves Open.

Maximum RCS Pressure Maximum S.G.

Pressure Unaffected

Loop, ADV Closed Auxiliary Feedwater Begins to Enter Affected S.G.

Safety Injection Actuation Signal Generated Safety Injection Punps Reach Full Speed

~

Set, int or Value 93+ of in'i,tial 4-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 Available ADV to Initiate Plant, Cooldown Operator Closes Affected ADV Block Valve Shutdown Cooling Initiated; RCS Average Temperature

~

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TABLE 4 Inte ated Steam Release LCSS OF AC AND STUCK OPEN ADV STEAM RELEASES Value Steam Release. Through Safety Valves Duri'ng 0 - 2 h.s, lhn Steam Release Through Atmospheric Steam Eh'alves During 0 - 2 hrs, ibm 4

Total Amount of Steam Released During 0'- 2 hrs, ibm Total Amount of Steam Released Until Shutdown Cooling is Initiated (325 F), ibm

]634]

742712 759053 913780

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