ML18227B253

06/08/1977 Letter Proposed Amendment to Appendix a of Facility Operating Licenses, as a Result of Re-Evaluation of ECCS Cooling Performance Calculated in Accordance with Approved Westinghouse Evaluation Model

ML18227B253
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
Download: ML18227B253 (58)

Text

. O. BOX 013100, MIAMI, FL 33101 gpss Ih~

s-6 ~ - <wx FLORIDA POWER & LIGHT COMPANY gegu1atory Docket Fg June 8, 1977 L-77-172 Office of Nuclear Reactor Regulation l1 Attention: Nr. Victor Stello, Director yQ Division of Operating Reactors U. S. Nuclear Regulatory Comm'sion Washington, D. C. 20555

Dear bi. Stello:

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CO Re: Turk y Point Units 3 and 4 Docket No. 50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 Zn accordance with 10 CFR 50.30, Florida Power 8 Light Company (FPL) submits herewith three (3) signed originals and, forty (40.) cop'es of a rectuest to amend Appendix A of Facility Operating Licenses DPR-31 and DPR-41.

This proposal is being submitted as a result of a =e-evaluation of ECCS cooling performance calculated in accordance with an approved Westinghouse Evaluation IÃodel. The proposed change is described below and shown on the accompanying Techn'cal Specification pages bearing the date of this letter in the lower right hand corner.

Pa e 3.2-3 Specification 3.2.6.a is revised sucn that the limit on the Heat Flux Hot Channel Factor fo- both Units 3 and 4 is reduced from 2.22(F~) to 2.20 for steam ge~ ator tube plugging in excess of lpga.

Pa es B3.2-4 and B3.2-6 Pages B3.2-4 and B3.2-6 present the basis for the revised li.'mit on Fq for botn Units 3 and 4.

771650242 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 & Light Company Nuclear Review Board. They have concluded that it does not involve an unreviewed safety question. A safety evaluation is attached.

l Very truly yours, Robert E. Uhrig Vice President REU/WAK/cmp Attachments cc: Mr. Norman C. Moseley, Region II Robert Lowenstein, Esquire

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reactivity, insertion u"". e.-ection greater than 0.3/ d k/k at eared pa- 'Q:aeperaale red vetth shall be determined withi.. -'ee'.-s.

b. A control rod shall be cc..s'"ared inoperable if (a) the rod cannot be moved by the CRD~f, or (b) the rod is misaligned x o its bank oy more than 15 inches, or (c) the rod drop time is not met.

c. If a control rod cannot be moved by the drive mechanism, shutdown margin sha11 be increased by boron addition to compensate for the wi'thdrawn worth of the, inoperable rod.

5 CONTROL ROD POSITION INDICATION If either the power range channel deviation alarm or the xod 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 alarms are inoperable for'wo hours or more,. the nuclear overpower trip shall be reset to 93X of rated power.

6. PONER DISTRIBUTION LIMITS

a. At all times except during, low power physics tests, the hot channel factors defined in the basis must meet the following limits:

F (Z) < (2.22/P) x K(Z) for P > .5 F (Z) < (4.44)*X K(Z) for P < .5 F~ < 1.55 fl + 0.2 (1-?))'here P is the fraction 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 t

Fq

/

~

• For tube plugging in excess of 10<, geese values'become (2 20/P) and (4.40) xespectively

b. Following initial loading before the reactor is operated above 75K of rated power and at regular effective full. rated power bimonthly intervals thereafter, power distribution maps, using the

'ovable detector system shall be made, to conform that the hot channel factor limits of the spe=ifica-tion are satisfied. For the purpose of this comparison, 3~2 3 6/8/77

l5 )

~ 0 An upper bound envelope of 2.22 cimes the nor I'==.'=aking factor axial dependence of Figure 3.2-3 has been determined =o .=-.= consistent =ith the technical specifications on po"er distribution c-n"=cl as given in Section 3.2.

%hen an F measurement is taken, both elope Mental er or and manu acturing tolerance must be allowed for. Five percent is the approp iate experimental uncert~~ ty allowance for a full core map taken with the movable incore detector flux mapping system aad three percent is the appropriate allowance for manufacturiag tolerance.

In the speci&ed li=~t o~f P, there is aa 8 percent allowance for uacerta ties which meaas that no=. ~ operation of the core is expected to result in 8~<1.35/1.08. De 1ogic behind the larger uncertainty in this case is that (a) normal pertur&ations 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 movemeat of rods, and can.limit it to the desired value, he has no direct control over ~~ and (c) an error ia the predictioas for radial power shape, which may be detected during startup physics tests can be compensated for in F by tighter axial control, q

but compensation for r~ is less readily available. %hen a measurement of F< ve is taken, experi ental error must be ~owed for and 4X is the appro-priate allowance for a full core map taken with the movable incore detector flux mappiag system.

Measurements of the hot channel factors are required as pa-t of start-up physics tests, at least. once each full rated power~ath of operation, and whenever zbaor~ power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The "core map taken followiag initial loa"ing provides confirmation of the basic nuclear'For steam generator tube plugging in excess. of 10$ , this value becomes 2.20.

B3.2-4

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Flux Difference (5g) and a reierence Value -h-ch --=asponds to tne full design power equilibrium value of Axial Offset (.-a:i=-'f set ~ <~/fractional

. power). The reference value of ~lux difference varies w=th power level and burnup but'xpressed. as axial offset it varies only =-- th burnup.

The technical specifications on power distr bution control assu a that the P upper bound envelope of 2-22 times Figure 3.2-3 is mt exceeded and xenon distributions are not developed which at a later time, ~auld cause greate local power peaking even though the flux differen'ce is then within the l~ts specified by the proce uze.

The target (or re ezence) value, of flux difference is determined as follows At any time that equilibrium xenon conditions have, been established, the i=

dicated flux dizfezence is noted .with part length .zods withdrawn from the co~

and'ith tne full length zod control rod bank more than 190 steps withdrawn'i.e.,

normal rated power operating pos'zion appropriate for the time in l=fa Control rods aze usually withdrawn farther as burnup proceeds). This value divided by th fraction of design power at which the core was operating is tae design power value of the taz"et flux difference. Ualues for all other core power levels are obtained by multiplying the design power value by tha fractional power. Since the indicated equilibrium value was noted, no allowances for excora detector error are necessary and indicated deviation of

+5Z hl ar permitted from the indicated reference va ue. During periods where extensive load fo lowing is zequired, it may oe impractical to estab~

the required core conditions for measuring 2e t~zt flux difference evez=

rated power month. For this reason, m thods are pe~tted by Item 6c of Section 3.2 for updating tha target 'flux differences. Figure B3 2-1 shows a typical construction of the target flux difference band at BOL and Figure B3.2-2, shows the typical variation of the full power value with burnup.

Strict control of the flux difference (and rod position) is not as necessary during part pow r operation. This is because. xeno'n-distribution control at part power is not as significant as the cortzol at full power. and allo~ance has been made in predicting the heat flux peaking factors for less strict co=-

trol at part pc-' Stzict control of the f1ux difference is not possible durga~ certai" = ..-:sics tests or during the required, periodic excore calibra-

• For steam generator tube plugging in, excess of 10%, this value becomes 2."0.

B3.2-6 6/8/77

ig 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. 15% steam generator tube plugging 2~ Fq = 2 20

3. 875 ft3 accumulator minimum water volume

4. 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.

ig 15 TABLE il LARGE BREAK TI51E SE UENCE OF EVENTS DECL (CD=0.4)

(Sec)

START 0.0 Rx Trip Signal 0.595 S. I. Signal 0.67 Acc. Injection 16.0 End of Bypass 27.16 End of Blowdown 27.31 Bottom of Core Recovery 46. 24 Acc. Empty 60. 8 Pump Injection 25.67,

i( 0 .

TABLE .'2

'ARGE BREAK DECL (CD=0.4)

Results Peak Clad Temp. F 2173 Peak Clad Location Ft. 6.5 Local Zr/H20 Reaction (max)X 11.655 Local Zr/H20 Location Ft. 6.0 Total Zr/H20 Reaction ~~ <0.3 Hot Rod Burst Time sec 22.6 Hot Rod Burst Location Ft. 6.0 Cal cul ati on Core Power Mwt 102~ of 2200 Peak Linear Power kw/ft 102~~'of 12.499 Peaking Factor 2.20 Accumulator Mater Volume (ft ) 876 (per accumu1ator)

Fuel region + cycle analyzed Cycle Region UNITS 3 5 4 3 3

15 TABLE LARGE BREAK CONTAINMENT OATA ORY CONTAINMENT NET FREE VOLUME 1.55x106 Ft3 INITIAL CONOITIOHS

Pressure 14.7 psia

~

Temperature 90 'F RMST Temperature 39 F Service Mater Temperature 63 F Outside Temperature 39 oF SPRAY SYSTEM Number of Pumps Operating 2 Runout Flow Rate 1450 gpm Actuation Time 26 secs SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating 3 Fastests Post Accident Initiation of Fan Coolers 26 secs

ii iQ TABLE 3 (Continued)

LARGE BREAK CONTAINMENT DATA DRY CONTAINMENT STRUCTURAL,HEAT SINKS Thickness In Area Ft )

Steel 0.03 31,400 Steel 0;063 107,158 Steel 0.1 56,371 Steel 0.2 57,185 Steel 0. 24 9,931 Steel 0.. 2898 Concrete 24.0 136,000 Steel 0.4896 23,677 Steel 0.6396 6,537 Steel 0.8904 4,915 Steel 1.256 27,802 Steel 1.56 5,307 Steel 2.0 668 Steel 2.?5 1268.7 Steel 5.5 1277.4 Steel 9.0 260. 4 Stainless 0.14 Concrete 24.0 14,392 Stainless 0.44 768 Stainless 2.126. 3,704 Stainless 0.007 102,400 Conc'te 24.0 59,132

il TABLE REFLQOD MASS AND BIERGY RELEASES FOR LIMITIHG BREAK DECLG CD = 0.4 Total Mass Flowrate Total Energy f1owrate Time, Sec LBm/Sec 10 BTU/Sec 46.235 0.0 0.0 48.36 0.0 0.0 53.982 35.09 0.4562 64'.197 93.67 1.16 76.997 96.36 1.20 92.197 117.2 1. 30 108.097 238.3 1.61 124.697, 267.3 1.65 160.397- 276.7 1.5?

199.397 283.3 1.48

0 Or 6.0 FT AHO 6.S FT 4 6 8 IO~ 2 6 8 lo' tf 6 & l02 2 luE (SEC)

Figure 1. Fluid Quality - DECLG (CD = 0.4)

0 0 6.5 ET 30 W

I 2p s.o rr ~

lp 60 F'f AtlD 65 FT

-Ip

-20 Ipl *2 IO' tI 6 8 lp 2 6 8 6 8 IO~ 2 6 8 lp~

TII.IE (SEC)

Figure 2. Mass Velocity DECLG (CD = OA)

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il

)0' 6

0 t

cv 2

)02 8

6 6.0 FT W

CD c/I 0)

)

8 I- 6.5 FT

)O0 0 ')

00 200 300 900 800 z)uE (sEc)

Figure 3. Heat Transfer Coefficient DECLG (Co = 0,4j

~

0 2500 2000 l500 g l000 500 0

0 10 20 30 ilo 50 Til<E (SEC)

Figure 4. Core Pressure - DECLG (CD = OA)

0 l.lxl05 SxlO" 7xlo" 5x I 0 lxl0 I 0 IO 20 30 tIO 50 T IME (SEC)

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

IO il 70 50 25 c5 0 ED O

-50

~ ~

. -70 0 IO 20 30 40 50 TiwE (sEc} .

Figure 6. Core f'ressure Drop DECLG (CD = OA)

~

Ih 2500 6.0 FT 2000 6.5 FT 6.5 FT 6.0 FT l500 co l 000 500 0

0 50 75 loO l25 150 200 TlhlE (SEC)

Figure 7. Peale Clad Temperature - DECLG (CD = OA)

il IS 2000 l750 l500 l250 l000 750 6.0 FT A}l0 6.5 FT 500 250 0

0 l00 200 3QQ ilQQ 500 VlVE (SEC)

Figure 8. Fluid Temperature - DECLG (CD = 0.4I

7000 5000 2500 YOP 0

I BOlTG~I

-2500

-5000

-7000 0 IO 20 25 30 TIME (SEC)

Figure 9. Coro Flow - Top and Bottom - DECLG (CD = OA)

15 20.0 l7.5 OOM(COMER LEVEL l5.0 l2.5 l0.0 75 CORE LEVEL 5.0 "2.5 0.0 0 25 50 75 l00 l25 l50 l75 200 TlWE (SEC)

Figure 10. Refiood Transient - DECLG (CD = OA) Downcomer and Core Water Levels

Il

2. 00 1.75 1.50 1.25 1.00

0. 75

0. 50

0. 25 50 100 150 175 200 TIWE (SEC3 Figure 10a. Ref lood Transient - DECLG (CD = OA) Core inlet Velocity

6000 5000

<l000 3000 Cl LLI 2000 l000 0

0 20. 0 30. 0 50. 0 TIME (SECI Figure 11. Accumulator Flow (Blowtlown) - DECLG f CD = OA)

8.0 6.0 I),0 2.0 0 ll0 80 I20 I60 200 2>l0 2&0 320 Ttwe (sec)

Fiy>re 12. Pumped ECCS Flow (ReEEood) - DECLG (Cp = OA)

30 25 C9 20 W

l5 lp C) 0 0 tlo 8O lZO I 60 200 2<lo 280 320 360 <l00 TlllE (SEC)

Figure 13. Containment Pressure - DECLG (CD = OA)

4 lk l.o 0.8 0 6 0.4

.0. 2 0

0 lp,p 20.0 30.0 <10.0 50. 0 TlVE (SEC)

Figure 14. Core Power Transient ~ DECLG (CD = 0.4)

ii 5.5 x I07 9.5 x l07 w 3.5 x I07

" 2.5 x l07

~ l.5 x l07 0.5 x l06

-0.5 x l06 0 lo.o 20.0 30. 0 50.0 Tlute (SEC)

Figure 15. 8reak Energy Released to Containment

0 1

I loo La O

I l000 C4 I

lt I

CC 900 CQ 800 700 600 500

%00 L1l Ch O

CD 300 l-UJ 200 O l 00 CD I00 200 300 TlME (SEC)

Figure 16. Containment Wall Condensing'Heat Transfer Coefficient

STATE OF FLORIDA )

)

COUNTY OF DADE )

Robert E. Uhr'g, being f'st 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 E. Uhrig Subscribed and sworn'to before me this day of,. //(gn~ 19 /7 w/~Xuun NOTARY PUBLIC, in c~:</

and for the'ounty of Dade, State of Florida Ny commission expires:

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