ML18227B253
| 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
gegu1atory Docket Fg
. O. BOX 013100, MIAMI, FL 33101 gpss Ih~
s- ~
6 - <wx FLORIDAPOWER & LIGHT COMPANY June 8, 1977 L-77-172 Office of Nuclear Reactor Regulation Attention:
Nr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Comm'sion Washington, D. C.
20555
Dear bi. Stello:
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 yQ l1
@c~+Q~~~ go<
CO Zn accordance with 10 Company (FPL) submits and, forty (40.) cop'es of Facility Operating CFR 50.30, Florida Power 8 Light herewith three (3) signed originals of a rectuest to amend Appendix A 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 (F~)
fo-both Units 3 and 4 is reduced from 2.22 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
I~
r' "I
I
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
Oi I
b.
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.
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 tolerance must be allowed uncert~~ ty allowance for taken, both elope Mental er or and manu acturing for.
Five percent is the approp iate experimental 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<
is taken, experi ental error must be ~owed for and 4X is the appro-ve 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
IO l
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 P
upper bound envelope of 2-22 times distributions are not developed which distr bution control assu a that the Figure 3.2-3 is mt exceeded and xenon 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 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 16.0 27.16 27.31
- 46. 24
- 60. 8 25.67,
i(
0
TABLE
.'2
'ARGE BREAK DECL (CD=0.4)
Results Peak Clad Temp.
F Peak Clad Location Ft.
Local Zr/H20 Reaction (max)X 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 Peaking Factor Accumulator Mater Volume (ft )
2200 12.499 2.20 876 (per accumu1ator)
Fuel region
+ cycle analyzed Cycle UNITS 3 5 4 3
Region 3
15
TABLE LARGE BREAK CONTAINMENT OATA ORY CONTAINMENT NET FREE VOLUME 1.55x106 Ft3 INITIALCONOITIOHS
- Pressure
~
Temperature RMST Temperature Service Mater Temperature Outside Temperature 14.7 psia 90
'F 39 F
63 F
39 oF 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
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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 Conc'te 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.?5 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
il
TABLE REFLQOD MASS AND BIERGY RELEASES FOR LIMITIHG BREAK DECLG CD = 0.4 Time, Sec Total Mass Flowrate LBm/Sec Total Energy f1owrate 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 1.5?
1.48
0
Or 6.0 FT AHO 6.S FT 4
6 8 IO~
2 6
8 lo' luE (SEC) tf 6
& l02 2
Figure 1. Fluid Quality - DECLG (CD = 0.4)
0 0
30 6.5 ET W
I 2p s.o rr ~
lp 60 F'f AtlD 65 FT
-Ip
-20 IO' tI 6
8 lp 2
6 8 Ipl
- 2 TII.IE (SEC) 6 8
IO~
2 6
8 lp~
Figure 2.
Mass Velocity ~ DECLG (CD = OA)
il
)0' 6
0 t
cv 2
)02 8
6 W
CD c/I
) 0) 8 I-6.0 FT 6.5 FT
)O0 0
') 00 200 z)uE (sEc) 300 900 800 Figure 3. Heat Transfer Coefficient
~ DECLG (Co = 0,4j
0
2500 2000 l500 g
l000 500 0
0 10 20 Til<E (SEC) 30 ilo 50 Figure 4. Core Pressure
- DECLG (CD = OA)
0
l.lxl05 SxlO" 7xlo" 5x I 0 lxl0 I 0
IO 20 T IME (SEC) 30 tIO 50 Figure 5. Break Flow Rate - DECLG (CD = OA)
IO il
70 50 25 c5 0
-50
~ ~
. -70 0
IO 20 TiwE (sEc}
30 40 50 Figure 6. Core f'ressure Drop
~ DECLG (CD = OA)
Ih
2500 2000 6.0 FT 6.5 FT l500 6.5 FT 6.0 FT co l000 500 0
0 50 75 loO TlhlE (SEC) l25 150 200 Figure 7.
Peale Clad Temperature
- DECLG (CD = OA)
il IS
2000 l750 l500 l250 l000 750 500 6.0 FT A}l0 6.5 FT 250 0
0 l00 200 VlVE (SEC) 3QQ ilQQ 500 Figure 8. Fluid Temperature
- DECLG (CD = 0.4I
7000 5000 2500 YOP 0
I BOlTG~I
-2500
-5000
-7000 0
IO TIME (SEC) 20 25 30 Figure 9. Coro Flow - Top and Bottom
- DECLG (CD = OA)
15
20.0 l7.5 l5.0 OOM(COMER LEVEL l2.5 l0.0 75 5.0 CORE LEVEL "2.5 0.0 0
25 50 75 l00 TlWE (SEC) l25 l50 l75 200 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 TIWE (SEC3 150 175 200 Figure 10a.
Reflood Transient - DECLG (CD = OA) Core inlet Velocity
6000 5000
<l000 3000 Cl LLI 2000 l000 0
0
- 20. 0 TIME (SECI
- 30. 0
- 50. 0 Figure 11. Accumulator Flow (Blowtlown) - DECLG fCD = OA)
8.0 6.0 I),0 2.0 0
ll0 80 I20 I60 Ttwe (sec) 200
. 2>l0 2&0 320 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 TlVE (SEC) 30.0
<10.0
- 50. 0 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 Tlute (SEC)
- 30. 0 50.0 Figure 15. 8reak Energy Released to Containment
0 1
I loo La O
I C4I lt I
CC CQ l000 900 800 700 600 500 L1l Ch O
CD l-UJ O
CD
%00 300 200 l 00 I00 TlME (SEC) 200 300 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/~Xuunc~:</
NOTARY PUBLIC, in and for the'ounty of Dade, State of Florida
~-.;~pc tu;iC ST>>Tg Ce FLOhel st VOCE c<+~'~st ernie rue'I Ny commission expires:
<=i.-.-'"-.nay avv~o ooaaeo ~
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