L-78-271, 08/16/1978 Letter Thermal & Hydraulic Safety Limits

08/16/1978 Letter Thermal & Hydraulic Safety Limits

ML18227D799
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: ML18227D799 (7)

Text

REGULATORY INFORNATION DISTRIBUTION SYSTEN (RIDS)

DISTRIBUTION FOR INCOMING MATERIAL 50-2 5i REC: STELLO V ORG: UHRIG R E DOCDATE: 08/ih/78 NRC FL PWR Sc LIGHT DATE RCVD: 08/22/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED

SUBJECT:

LTR 3 ENCL 3 FURNISHING CLARIFI'CATION TO APPLICANT"S 07/10/78 PROPOSED TECH SPEC CHANGES RE ALLOWING FULL PWR OPERATION WITH UP TO 25/ OF THE STEAN GENERATOR TUBES PLUGGED> PERTAINING TO THE REACTOR CORE THERNAL HYDRAULIC LINITS... W/ATT SUPPORTING FIGURES.

PLANT MANE: TURKEY PT 03 REVIEWER INITIAL: X JN TURKEY PT 84 DISTRIBUTER INITIAL:QL

+++>+++++++4++44+ DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.

<DISTRIBUTION CODE A001)

FOR ACTION: BR CH IEF ORBOi BC+4 W/7 ENCL INTERNAL: EG FIL +W/ENCL NRC PDR4+W/ENCL

+W/2 ENCL OELD+4LTR ONLY HANAUER+<W/ENCL CORE PERFORNANCE BR++W/ENCL AD FOR SYS 8c PROJ++W/ENCL ENGINEERING BR+%M/ENCL REACTOR SAFETY BR++W/ENCL PLANT SYSTENS BR+4W/ENCL EEB++W/ENCL EFFLUENT TREAT SYS%+W/ENCL J. NCGOUGH44W/ENCL EXTERNAL: LPDR S MlAMIi FL++W/ENCL TERA~~W/ENCL .

NS IC++W/ENCL ACRS CAT B4+W/16 ENCL DISTRIBUTION: LTR 40 ENCL 39 C ONTROL NBR: 782340i72 8 I ZE: 2P+2P.

0-x %%%4 4%4 4 w%%4%%%4%44%%%%4%%4%%%%%% THE END w%4%%%%%%4 0%%%%%4 4%%%%%4%4 cpP

'C I I

RECIjLATB - Y DOCKET FfLE I C('.

P. O. BOX 529100, MIAMI.FL 33152 FLORIDA POWER 5 LIGHT COMPANY 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'0-251 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'o allow full power 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.l-lb, which are to be added to the Technical. Specifications to permit operation, with steam generator tube plugging between 19% and 25%.

The basic method's 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 ll surrounding fuel rods was modeled by means of the COBRA. ZXXC 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 Zor lower power levels: FPH was taken to be the highest, value permitted by the Technical Specifications, F~H = 1.55[1 + 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.

82340 ) 72

Mr. Victor Stello Page 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. Flow was reduced by another 1% to account for flow redistribution and the- effects of interassembly mixing. To 'generate Figure 2.l-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'xit 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 .1-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

FIGURE 1

.SUBCHANNEL GEOMETRY r'

~ ~ ~

, ir ~ d1 Wt ~

650 SATURATION 640 LINES 2400 630 psia 620 psia 610 .> 2100 psia 600

~1900 590 psia DNB LIMIT LINES 570 560 550 540 20 40 60 80 100 120 140 RATED POWER (PERCENT)

Figure 2 Reactor Core Thermal and Hydraulic Safety Limits