ML18227B286
| ML18227B286 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 08/16/1978 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| L-78-271 | |
| Download: ML18227B286 (10) | |
Text
REGULATORY INFORMATION.DISTRIBUTXON SYSTEM (AIDS)
DISTRIBUTXON FOR INCOMING MATERIAL 0-2 51 REC:
STELLO V
NRC ORG:
UHRIG R E DOCDATE: 08/16/78 FL PWR 8( LIGHT DATE RCVD: 08/22/78 DOCTYPE:
LETTER NOTARIZED:
NO COPIES RECEIVED
SUBJECT:
LTR 3 ENCL 3 FURNISHING CLARIFICATION TO APPLICANT"S 07/10/78 PROPOSED TECH SPEC CHANGES RE ALLOWING FULL PWR OPERATION WITH UP TO 25/
OF TflE STEAM GENERATOR TUBES PLUGGED>
PERTAINING TO THE REACTOR CORE THERMAL HYDRAULIC LIMITS... W/ATT SUPPORTING FIGURES.
PLANT NAME: TURKEY 'PT03 TURKEY PT 04 REVIEWER INITIAL:
XJM DISTRIBUTER INITIAL:,P4 DISTRIBUTION OF THIS 'MATERIAL IS AS FOLLOWS GENERAL, DISTRIBUTION FOR'FTER ISSUANCE OF OPERATING LICENSE.
<DISTRIBUTION CODE A001)
.'.FOR ACTION'NTERNAL':
DR CHXEF ORB51 BC4~W/7 ENCL REG FI
- wW/ENCL E++W/2 ENCL HANAUER+%W/ENCL AD FOR. SYS 5 PROJ+4W/ENCL REACTOR SAFETY 'BR%%W/ENCL EEB>+W/ENCL J;
MCGOUGH+<W/ENCL NRC PDR4+W/ENCL GELD>%LTR ONLY CORE PERFORMANCE BR<+W/ENCL ENGINEERING BR+<W/ENCL PLANT SYSTEMS BR+<W/ENCL EFFLUENT TREAT SYS>>W/ENCL EXTERNAL:
LPDR S MIAMIt FL~>+W/ENCL TERA~~W/ENCL NSIC++W/ENCL
'CRS CAT. B~.~W/16 ENCL DISTRIBUTION:
LTR 40 ENCL 39 CONTROL NBR:
782340172 SIZE:
2P+2P
ii P
4
P, O. BOX 529100, MIAMI,F L 33152 FLORIDAPOWER & LIGHTCOMPANY August 16, 1978 L-78-271 Office of Nuclear Reactor Regulation Attention:
Mr. Victor Stello, Jr., Director Division of Operating, Reactors U.
S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stello:
Re:
Turkey Point Units 3 and 4
Docket Nos.
50-250 and 50-251 Thermal and H draulic Safet Limits Florida Power and Light Company (FPL) letter L-78-230 of July 10, 1978 contained proposed changes to the Technical Specifi-cations of Turkey Point Units 3 and 4 to allow fullpower operation with up to 25% of the steam generator tubes plugged.
The staff has asked FPL to clarify thederivation of the Reactor Core Thermal Hydraulic Limits, Figure 2.1-lb, which are to be added to the Technical Specifications to permit operation with steam generator tube plugging between 19% and 25%.
The basic methods used to produce this figure are described in FPL report NAD-QR-25, submitted with letter L-77-106 on April 4,
1977.
A subchannel of the fuel assembly consisting of a RCCA guide thimble and 11 surrounding fuel rods was modeled by means of the COBRA IIIC thermal-hydraulics code; the geometry is shown in Figure l.
Of the eight subchannels, channels one and five contain an allowance for a water gap between adjacent assemblies.
A number of COBRA computer runs were made withNthe following input.
For power levels 100% of rated the FgH for each fuel rod was 1.55, while for lower power levels F$ H was taken to be the highest value permitted by the Technical Specifications, F~H = 1.55tl + 0.2(1-P)]
where P is the fractional power.
The axial power distribution for all fuel rods was a center peaked chopped cosine with a peak to average ratio of 1.55.
Fuel densification was accounted for by a reduction in active fuel length from 144 to 142 inches.
82340l72 PEOPLE..
~ SERVING PEOPLE
Ik
Mr. Victor Ste~
gage Two August 10, 1978 Cross flow mixing parameters and flow. correlations were taken to be those listed in Table 4 of NAD-QR-25.
In accordance with the methods described in that report the core coolant average mass velocity of 2..32
- 10 lb/hr ft.
was reduced by 5% to allow for inlet flow maldistribution.
Plow was reduced by another l% to account for flow redistribution and the effects of interassembly mixing.
To generate Figure 2.1-lb the coolant flow was further reduced by another 5% to account for the effects of the 25% steam generator tube, plugging.
Appropriate uncertainty factors were included for power levels, pressures and temperatures.
The curves of Figure 2.l-lb, covering a range of pressure
- levels, were generated by varying the average coolant temperature for a given power level until the minimum DNBR, as calculated with the COBRA model described
- above, was 1..24 with the rod bow penalty included.
Thus each curve represents the locus of points of the DNBR limit.
The DNBR was calculated with the W-3 correlation and the "L" grid correction for LOPAR 15xl5 Westinghouse fuel.
The DNBRs obtained from the COBRA computer runs were reduced by 28.9% to account. for the rod bow penalty prescribed for fuel with the highest burnup.
At the lower power levels the average coolant temperature at the reactor exit reaches the saturation temperature before the DNB limit is reached.
The saturation line then represents the limiting condition.
Figure 2 shows these saturation limit curves as well as the DNB limit curves.
To be consistent with the Reactor Core Thermal and Hydraulic Safety Limit curves presently in the Technical Specifications and those submitted for less than 25% tube plugging on June 22, 1978 (letter L-78-217), horizontal limit lines are conservatively prescribed in Figure 2.l-lb instead of the higher saturation limit lines.
Very truly yours, Robert E. Uhrig Vice President REU:RDH:lc cc:
Mr. Robert, Lowenstein, Esquire Mr. James P. O'Reilly, Region II
il,
, ~
I
FIGURE 1 SUBCHANNEL GEOMETRY t 3Q 7
I C
e e wvi 2) l I ~
c
iO 0
<4
650 640 SATURATION
/
LINES 630 2400 psia 0
00
+
0A 620 610 600 psza
~ 2100 psia 590
~1900 psia 580 DNB LIMIT LINES 570 560 550 540 0
20 Figure 2.
40 60 80 100 120 RATED POWER (PERCENT)
Reactor Core The'rmal and Hydraulic Safety Limits 140
i>
Il I
t l
V