L-77-172, Request to Amend Appendix a of Facility Operating Licenses. Proposal Submitted as Result of Re-evaluation of ECCS Cooling Performance Calculated in Accordance with Approved Westinghouse Evaluation Model

Request to Amend Appendix a of Facility Operating Licenses. Proposal Submitted as Result of Re-evaluation of ECCS Cooling Performance Calculated in Accordance with Approved Westinghouse Evaluation Model

ML18227D839
Person / Time
Site:	Turkey Point
Issue date:	06/08/1977
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-77-172
Download: ML18227D839 (38)

Text

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Mr. Victor Stello'ROM:

Florida Power

& Light:.Company

~ Miami, Fla Mr. Robert ED Uhrig DATC Of" DOCUt'ICIfT 6/8/77 DATE nCCEIVCD 6/13/77 ETTEn

~oronfzED henIOINAL.."'>>

~NcL'ASSIFIED, CfcoPv DEScnIPTION pnop l

,INPUT FoflM ENCLOSUfiE Amdt, to-OI/change to specs'~

,sunmitted as re-evaluation of ECCS

. notorized 6/8/77 '

~ ~ ~

IQUMOEn OF COPIES fIECEIVCD gc g~C~

lf "II Appendix A tech I

a result of a cooling performance.,;

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FLORIDAPOWER & LIGHTCOMPANY l!

June 8, 1977 L-77-172 Office of Nuclear Reactor Regulation Attention:

Mr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.

20555

Dear Mr. Stello:

AO II (0

'ggl

~$7 u.>.

Re:

Turkey Point, Units 3 and 4

Docket No. 50-250 and 50-251 Proposed Amendment to Facility O~eratin Licenses DPR-31 and DPR-41 In accordance with 10 Company (FPL) submits and forty (40) copies of Facility Operating CFR 50.30, Florida Power

& Light herewith three (3) signed originals of a request to amend Append'x A Licenses DPR-31 and DPR-41.

This proposal is being submitted as a result of a re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model.

The proposed change is described below and shown on the accompanying Technical Specification pages bearing the date of th's letter in the lower right hand. corner.

Pa e 3.2-3 Specification 3.2.6.a is revised such tnat the limit on the Heat Flux Hot Channel Factor (Fq) for both Units 3 and 4 is reduced from 2.22 to

2. 20 for steam ger.gator tube plugging in excess of 10%.

Pa es B3.2-4 and B3.2-6 Pages B3.2-4 and B3.2-6 present the basis fo" the revised limit on Fq for both Units 3 and 4.

PEOPLE... SERVING PEOPLE

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Office of Nuclear Reactor Regulation

'Attention:

Mr. Victor Stello, Director Page Two The proposed amendment has been reviewed by the Turkey Point.

Plant Nuclear Safety Committee and the Florida Power 6 Light Company Nuclear Review Board.

They have concluded that it does not involve an unreviewed safety question.

A safety evaluation is attached.

Very truly yours,

,~g.-C'~

g/l'obert E. Uhrig Vice President REU/WAK/cmp Attachments I

cc:

Mr. Norman C. Moseley, Region II Robert Lowenstein, Esquire

reactivity insertion u~

.-. =-'.ection greater than 0.3Z 4 k/k at rated po.eQ:oopeeahle rod worth shall be determined withitn -'eeks.

b.

A control zod shall be cons="ezed inoperable if (a) the zod cannot be moved by the CRT/, or (b) the rod is misaligned zro its ban'c by more than 15 inches, o"

(c) the rod drop time is not net.

c. If a contzol rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addition to compensate f'r the withdrawn worth of the. inoperable rod.

5.

CONTROL ROD POSITIO'.I INDICATION If either the power range channel deviation alarm or 1

the rod deviation monitor alarm are not operable rod

~ positions shall be logged once per shift and after a load change greater than 10% of rated power. If both

'larm are inoperable for'wo hours or moze,. the nuclear overpower trip shall be reset to 93% of rated power.

6.

POLER DISTRIBUTIO'8 LIMITS a.

At all times except during low power physics tests, th'e hot channel factors defined in the basis must e

meet the following limits:

F (Z) < (2.22/P) x K(Z) zor P >.5 F (Z) < (4.44) k K(Z) for 'P <.5 q

F~

< 1.55 fl + 0 2 (1-P)1 where P is the'raction of rated power at which the core is operating.

K(Z) is the function given 'in Figure 3.2-3 and Z is the core height location of Fg b.

For tube plugging in excess of 10@,

values become (2.20/P) and. (4.40) respectively.

Following initial loading before the reactor is operated above 75% of rated power and at regular effective full rated power monthly intervals.,

thereafter, power distribution maps, using the

'ovable detector system shall be made, to conform that the hot channel factor limits of the specifica-tion are satis ied.

For the purpose of this comparison, t

3 ~ 2 3 6/8/77

A An upper bound envelope of 2.22 "times the nor. I'==-'eeking factor axial dependence of Figure 3.2-3 has been determined "o

'=-.= consistent

-~ith the

. technical specifications on power distribution c-n =ol as given in Section 3.2.

,Vhen an F measurement is taken, both ezpe ~ental error and manufacturing tolerance must be allowed for.

Five percent is the appropriate experimental uncertainty allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

In the specified li=~t o" P>H, there is an 8 percent allowance for uncertz~-

ties which means that no-. mal operation of the core is expected to result in P~(1.55/1.08.

~~e logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalign-ment) affect F><, in most cases without necessarily affecting F, (b) the operator has a direct influence on F through movement of rods, and can.limit q

it to the desired value, he has no direct control over P~ and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests can be compensated for in F by tighter 'axial, control, g

but compensation for r~ is less readily available.

%hen a measurement of P>

ks taken, experimental error must be allowed for and 4X is the appro-priate allowance for a full core map taken with tne movable incore detector, flux mapping system.

Measurements of the hot channel factors are required as part o

start-up physics tests, at least. once each full rated powermnth of operation, and whenever abnormal power dist'ribution conditions require a reduction of core power to a level based on measured hot channel factors.

The '"core map taken following initial loading provides confirmation of'he basic nuclear

• For steam generator tube plugging in excess of 10%, this value becomes 2.20.

B3.2-4

.6/8/77

Flux Difference (hg) and a reference value which c-=esponds to the fu13.

design power equilibrium value of Axial Offset

(.-'a:=.=

0 set

= 'r/fractional

, power).

The reference value of flux difference var'es w th pow r level and burnup but expressed as axial offset it varies only."th burnup.

The technical specifications on power distr.bution ccntzol assure that the F

upper bound envelope of 2.22 times Figure 3.2-3 is not exceeded and xenon distributions are not develooed which at a later tine, would cause greater local power peaking even though the flux differen'ce is tnen within the l&.ts specified by the procedure.

The target (or reference) value of flux difference is determined as fo13.ows At any time that equilibrium xenon conditions have been established tne i=

dicated flux dizference is noted.wit.x part length rods withdrawn from tne co~~

and with tne full lengtn rod control rod bank, more than 190 steps withdrawn'i.e.,

normal rated power operating posi.tion appropriate for the time in 1" fa Control rods are usually withdrawn farther as burnup proceeds).

This value divided by the fraction of design power at which the core was operating is tne design power value of the tar et flux difference Va3.ues for all other cora power levels are obtained by multiplying tne design power value by the fractional power.

Since the indicated equilibrium valu was noted, no allowances for excore detector error ar necessary and indicated deviation of

+5K DI are permitted from the indicated reference value.

During periods where extensive load fo3.lowing is required, it may be impractical to'stab~

the required core conditions for measuring the t~~ t flux difference every rated power month.

For this reason, methods aze pe~tted by Item 6c of Section 3.2 for updating the target flux differences.

Fibre B3.2-1 shows a

typi.cal construction of the target flux difference band at. BOL and Figure B3 ~ 2-2 shows the typical variation of the full. power value wi.th buznup.

Strict control of the flux difference (and rod position) is -ot as necessary during part pow r operation.

This is because xenon. distribution control at part power is not as significant as the control at full power'nd allowance has been made in predicting t'se heat flux peaking factors for less strict co"..

trol at part pc=':.

Strict control of the flux difference is not possible during certai=

"-.'sics tests'or during the

/

• Fpz stear. generator tube plugging becomes 2."0.

az.2-6 required, periodic excore calibra'-

in excess of 10%, this value 6/8/77

SAFETY EVALUATION I.

Introduction This safety evaluation and the attached Westinghouse ECCS re-evaluation support the following proposed change to the Technical Specifications:

(1)

The maximum allowable nuclear peaking factor (Pq) is decreased from 2.22 to 2.20, for steam generator tube plugging in excess of 10%.

II.

Discussion A re-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation Model has been performed.

The re-evaluation shows that for breaks up to and including the double ended severence of a reactor coolant pipe, the ECCS will meet the Acceptance Criteria presented in 10 CFR 50.46.

The detailed re-evaluation is

attached, and shows that, at a core power level of 102%

of 2200 Mwt and a minimum accumulator water volume of 875 ft3 per accumulator, the maximum allowable nuclear peaking factor is 2. 20 for steam generator tube plugging in excess of 10%.

The attached Westinghouse ECCS re-evaluation assumed:

1.

2.

3.

4

~

15% steam generator tube plugging Fq = 2.20 875 ft3 accumulator minimum water volume 2200 Mwt core power level III.

Conclusions Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and does not reduce the margin of safety as defined in the basis for any technical specification; therefore, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and (3) such activities will be conducted in com-pliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

LARGE BREAK TIME SE UENCE OF E'/ENTS DECL (CD=0.4)

(Sec)

START Rx Trip Signal S. I. Signal Acc. Injection End of Bypass End of Blowdown Bottom of Core Recovery.

Acc.

Empty Pump Injection 0.0 0.595 0.

67'6.

0 27.16 27.31

46. 24

60. 8 25.67

TABLE LARGE BREAK DECL (CD=0.4)

Results Peak Clad Temp. 'F Peak Clad Location Ft.

Local Zr/H20 Reaction (max)%

Local Zr/H20 Location Ft.

Total Zr/H20 Reaction Hot Rod Burst Time sec" Hot Rod Burst Location Ft, 2173 6.5" 11.655 6.0

<0. 3 22.6 6.0 Cal cul ati on Core Power Mwt 102% of Peak Linear Power kw/ft 102%'of

'eaking Factor Accumulator Water Volume (ft )

2200 12.499 2.20 875 (per accumulator)

Fuel region

+ cycle analyzed Cycle UNITS 3 5 4 3

Region 3

TABLE 3

LARGE BREAK CONTAINMENT OATA ORY CONTAINMENT NET FREE VOLUME 1.55x10 Ft INITIAL CONOITIONS Pressure Temperature RNST Temperature Service Mater Temperature Outside Temperature 14.7 psia

.90 F

39 OF 63 oF 39

'F SPRAY SYSTEM Number of Pumps Operating Runout Flow Rate Actuation Time 2

1450 gpm 26 secs SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating Fastests Post Accident Initiation of Fan Coolers 3

26 secs

TABLE 3 (Continued)

LARGE BREAK CONTAINMENT DATA DRY CONTAINMENT STRUCTURAL HEAT SINKS Thickness In Area Ft Steel Steel Steel Steel Steel Steel Concrete Steel Steel Steel Steel Steel Steel Steel Steel Steel Stainless Concrete Stainless Stainless Stainless Concrete 0.03 0; 063 0.1 0.2

0. 24-0.2898 24.0 0.4896 0.6396 0.8904 1.256 1.56 2.0

2. 75 5.5 9.0 0.14 24.0

0. 44

2. 126.

0.007 24.0 31,400 107,158 56,371 57,185 9.931 136,000 23,677 6,537

, 4,915 27,802 5,307 668 1268. 7 1277.4 260. 4 14,392 768 3,704.

102,400 59,132

TABLE I4 REFLOOD tNSS AND ENERGY RELEASES FOR LIMITING BREAK DECLG CD = 0.4 Time, Sec Total Mass Flowrate LBm/Sec Total Energy Flowrate 10 BTU/Sec 46.235 48.36 53.982 64.197 76.997 92.197 108.097 124.697 160. 397 199. 397 0.0 0.0 35.09 93.67 96.36 117.2 238.3 267.3 276.7 283.3 0.0 0.0 0.4562 1.16

1. 20

1. 30 1.61 1.65 1.57 1.48

g 0.95 6.0 FT A}lD 6.5 FT

0. 90 I

0.85 0.80

~ ~

0.75 IO-'

8 l00 6 slo' T IIIE (SEC) 6 8 IO 2

'll 6

8 IO'"

~ Figure 1. Fluid Quality - DECLG {CD = OA)

tip 30

~

6.5 IT I

2P 6.0 IT ED UJ) lo 6-0 FT AND 6.5 FT

- IO

>>20 6

8 lp 2

6 8 )Pl 2

TIVE (SEC) 6 8 lp 2

6 8

lp'igure

2. Mass Velocity - DECLG (CD = 0,4)

i 03 8

6 U

I cv 2

I-

)02 8

6 2

6.0 FT CO

)0)

I-l-6.5 FT 2

lo0 0

f00 200 TiVE (SEC}

300 900 500 Figure 3.

Ideat Transfer Coefficient - DECLG (CD = 0.4)

2500 2000 l500 l000 500 0

IO 20 TIME (SEC) 30

'IO 50 Figure 4. Core Pressure

- DECLG (CD = 0.4)

I.lxlQ5 exlo" 7xlo" 5xlQ" i5 3xlo" lxlQ"

-lxlo" 0

IQ 20 TIME (SEC) 30'IQ 50

~

Figure 5. Break Flow Rate - DECLG (CD = OA)

70 50 c5 0

-50

-70 0

10 20 TlME (SEC) 30 00 50 Figure 6. Core Pressure Drop - DECLG (CD = OA)

2500 2000 6.0 FT 6.5 FT l500 ED 6.5 FT 6.0 FT cB

!000 O

C)

CD 500 0

0.

25 50 75 l00 TlhlE (SEC) l 25 l50 l75 200 Figure 7, Peals Clad Temperature

- DECLG (CD = 0.4)

2000 1750 1500 U

1250 oc 1000 750

$ ~

250 6.0 FT AND 6.5 FT 0

l00 200 T111E (SEC) 300 900 500 Figure 8. Fluid Temperature

- DECLG (CD = 0,4)

7000 5000 2500 TOP 0

\\

MI'IOM

-2500

-5000

<<7000 0

l0 l5 TlME (SEC) 20 30 Figure 9. Core Flow - Top and Bottom - OECLG (CD = OA)

20.0 17.5 15.0 OONHCOHER LEVEL

12. 5 10.0 7.5 5.0 CORE LEVEL "2.5 0.0 0

25 50 75 100 TlhIE (SEC) 125 150 175 200 Figure 10.

Reflood Transient

- DECLG (CD = OA) Downcomer and Core Water Levels

2.00 l.75 I.50 o

0,75

0. 50

0. 25 0.00 50 l00 T-lltE {SEC) l25 l 50 Figure 10a.

Reflood Transient'- DECLG (CD = OA) Core inlet Velocity

6000 5000 CD UJ 4000 3000.

Ch 2000 CD l000

~t 0

0 lo.o

20. 0 TlME (SEe) 30.0

'00. 0 50.0 Figure 11. Accumulator Flow (Blowdown) - DECLG (CD = OA)

8.0 o

6.0 Q.O 2.0 0

80

) 20 l60 T)ME (sec) 200 290'80 320 Figure 12. Pumped ECCS Flow (Reflood) - DECLG (CD = OA)

30 25 CV 20 ul c

l5 I-IO

~cfI-CD 0

0 lIO 80 I20 IGO 200 '<)0 280 320 360

. TINE (SEC)

Figure 13. Containment Pressure

- DECLG (CD = 0.4)

l.p 0.8 p

K p g 0.2 0

lp.p 20.0 TIME (SEC) 30.0 50.0 Figure 14. Core Power Transient

- DECLG (CD = 0.4)

3.5 x l07 2.5 x l07 l.5 x 107 0,5 x I06

-0.5 x I06 0

I0.0

20. 0 TIME (SEC) 30.0 50.0 Figure 15. Break Energy Released to Containment

I I 00 l000 900 800

/

700 600 500

'400 300 200 IOO 0

0 IOO TIME (SEC) 200 300 Figure 16. Containment Wall Condensing Heat Transfer Coefficient

STATE OF FLORIDA

)

)

COUNTY OF DADE

)

ss Robert E. Uhrig, being first duly sworn, deposes and says:

That he is a Vice President of Florida Power

& Light Company, the Licensee herein; That he has executed the foregoing document; that the state-ments made in this said document.

are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.

Robert ED Uhrig Subscribed and sworn to before me this day of p/,~'M 19 77

~Spud ~ 6/,/

NOTARY PUBLIC, in and fbr th) County of Dade, State of Florida 8"Th 'C l'0"UC 5$ATg Of FLOftr A st MFC~

UY CEAAtdSlW tg1CS HAY 5,

NS1 Ny commission expires'a

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