ML20209D039

Demonstration of Conformance of Jm Farley Units to App K & 10CFR50.46 for Large Break Locas

ML20209D039
Person / Time
Site:	Farley
Issue date:	06/30/1986
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20209D006	List: ML20209D001 ML20209D026 ML20209D039
References
TAC-62283, TAC-62284, NUDOCS 8609090172
Download: ML20209D039 (84)

Text

.

ATTACHMENT 3

~

DEMONSTRATION OF THE CONFORMANCE OF JOSEPH M. FARLEY UNITS TO APPENDIX K AND 10CFR50.46 FOR IARGE BREAK LOCAs Westinghouse Electric Corporation Nuclear Technology Division Nuclear Safety Department Safeguards Engineering and Development June 1986 t

8609090172 860825 PDR ADOCK 05000348 P PDR

-- .-. - - _ . . . - . - . - . _ _ _ _ . - _ - = - _- - -- --

I. Introduction This document reports the results of an analysis that was performed to i drmonstrate that Joseph M. Farley, Units I and II, meet the requirements of Appendix K and 10CFR50.46 for Large Break Loss-of-Coolant-Accidents 1

(LOCA).

l II. Method of Analysis i

i Tho analysis was performed using the Westinghouse 1981 Evaluation Model with BART for a spectrum of break coefficients. The Westinghouse 1981 ECCS Large Break Evaluation Model was developed to determine the RCS

rocponse to design basis large break LOCAs (see References 6-13,19-20).

Tho hydraulic analyses and core thermal transient analyses were performed with the 1981 Evaluation Model code using 102 percent of licensed NSSS i j coro power. The 1981 Evaluation Model is comprised of the SATAN-VI, j HREFLOOD, COCO, and LOCTA-IV computer codes (References 2,3,4 and 1, roopectively, see also Reference 8). When the BART code (References I 16-18) is used along with the 81EM, slightly different versions of the l HREFLOOD (Interim Hreflood) and LOCTA (Interim LOCTA) are required. The

! purposes and structures of these codes do not differ significantly from l tho standard 81EM Model. The SATAN-VI code was used to generate the i blowdown portion of the transient, the HREFLOOD CODE was used to generate

! tho refill /reflood system hydraulics, and the COCO code was used to Ovaluate the containment response. The BART code is based primarily on initial conditions taken from the results of Interim EREFLOOD and Interim

. LOCTA at Bottom-Of-Core recovery (BOC time) and on the reflooding rate.

l BART provides an improved system hydraulic calculation. Cladding thermal i analyses were performed with the LOCTA-IV code which uses the RCS

procsure, fuel rod power history, steam flow past the uncovered part of tho core, and mixture height history from the SATAN-VI, WREFLOOD, and BART codes as input.

l Tho iuel parameters used as input for the LOCA analysis were generated using the Revised PAD Thermal Safety Model (References 14-15). The Large Broak LOCA analysis for Joseph M. Farley was performed at a power level of 2652 MWt. Other pertinent assumptions include a 10% steam generator tubo plugging level, 17 X 17 STD fuel design which is the current design for both Farley units, and an upflow barrel-baffle configuration. The upflow barrel-baffle configuration has previously been shown to represent c omall peak clad temperature penalty, hence the use of this configuration is conservative and bounding on both units. This analysis c100 incorporates a conservatively small total reactor coolant system ficw (14 below tech spec limit).

l i

i

J Tchle 1 shows the time sequence of events for the Large Break LOCA

'

• transients. Table 2 provides a brief summary of the important results of tho LOCA analyses for each case. Figures 1 and 2 show important core 3 characteristics during the blowdown phase of the transient (Core Pressure Cnd Core Flow versus Time, respectively). Figures 3 and 4 indicate the ficw of ECCS water into the RCS (Accumulator Flow and Pumped ECCS Flow i

a during Reflood versus Time, respectively). The flooding rate during the raflood portion of the transient are given in Figure 5. Clad Average Tcaparatures as a function of time indicating peak clad temperatures are given in Figure 6.

Three break size coefficients were evaluated;Co = 0.4, CD = 0.6, and Cp = 0.8. These transients were considered to be terminated when the

hat rod clad average temperature " turned around" (i.e. - hot rod clad cvarage temperature began to decline) indicating that the peak clad j tcaparature had been reached.

III. Results of the three break sizes evaluated, the CD= 0.4 break proved to be the

-liniting (highest PCT) case with 0 a peak 0clad temperature of 1973 F, 0

, ccupared with PCTs of 1787 F and 1635 F for the CD = 0.6 and CD = 0.8 cases, respectively.

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IV. Conclusions Fcr breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptance criteria as presented in 10CFR50.46. That is:

1. The calculated peak clad temperature does not exceed 2200 0 F based on a total peaking factor of 2.32.

2. The amount of fuel alement cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircalloy in the reactor.

3. The clhd temperature transient is terminated at a time when the core geometry is still amenable to cooling.

4. The cladding oxidaticn limits of 17% are not exceeded during or after quenching.

5. The core temperature is reduced and the decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

TABLE 1 IARGE BREAK TIME SEQUENCE OF EVENTS DECLG (CD = 0.8) DECLG (CD = 0.6) DECLG (CD = 0.4)*

(Sec) (Sec) (Sec)

Stort 0.0 0.0 0.0 R00ctor Trip 0.497 0.503 0.511 Signal S.I. Signal 0.690 0.780 0.95 Acc. Injection 9.38 11.50 15.30 Pump Injection 25.690 25.780 25.95 End of Blowdown 21.644 24.854 31.515 Cottom of Core 34.401 37.604 44.40 R3covery Acc. Empty 43.985 47.053 52.404

• Limiting Break d

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TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS RESULTS DECLG (CD = 0.8) DECLG (CD = 0.6) DECLG (CD = 0.4)

P0nk Clad Temp., OF 1635. 1787. 1973.

Pack Clad Temp. 7.0 7.25 7.25 Location, ft Local Zr/ 1.193 1.860 3.219 H 2 O Rxn (max), 4 Local Zr/ 7.25 7.25 6.0 H 2 O Location, ft Tctal Zr/H 2 O Rxn, 4 <0.3 <0.3 <0.3

, Hot Rod Burst Time, 90.4 68.4 41.0 COC Hot Rod Burst 7.25 6.5 6.0 Location, ft INPUT NSSS Power, MWt, 102% of 2652 P0ck Linear Power, kw/ft, 102% of 12.07 P02 king Factor (At Design Rating) 2.32 Accumulator Water Volume (Cubic i Fcet per Tank) 1025.

Accumulator Pressure, psia 600 Number of Safety Injection Pumps Cperating 2 StOtm Generator Tubes Plugged 10% (uniform),

• Uniform 10% Steam Generator Tube Plugging assumes 10% SG tubes plugged in each steam generator and corresponds to the worst plugging level in any steam generator.

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