ML20199G187
| ML20199G187 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/19/1997 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19313D093 | List: |
| References | |
| NSD-SAE-ESI-97, NSD-SAE-ESI-97-647, NUDOCS 9711250116 | |
| Download: ML20199G187 (24) | |
Text
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NSD-SAE-ESI-97-647 SNC RESPONSE TO NRC RAI ON BELOCA November,1997
% estinghouse Electric Corpration Energy System Business Unit ;
P.O. Ikw 355 Pittsburgh, PA 15230 4 355 l
01997 Westinghouse Electric Corporation l All Rights Reserved 9711250116 971119 ^
PDR P ADOCK 05000348 pg
- 12. The plant spectffc modeling and analysis for the large break loss of coolant accident (LOCA) is not provided. Please provide additional information regarding the analysis assumptions for the best estimate large bruk LOCA calcuir .ms. Information needed is the assumed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, singlefailure assumptions, loss-of offsite power assumptions.
time step assumptions, and major plant parameters with uncenainties. Show that the calculations were performed with the approved version of WCOBRA/ TRAC revision I, andprovide information that shows compliance with the code limitations and restrictions.
(WCAP 14723. Section 6 3).
Response to RAI-12:
A summary of results of the Best Estimate (BE) LBLCCA analysis performed for Farley Units and 2 power uprate using ECOBRA/ TRAC was provided in WCAP 14723, Section 6.1.1 (Reference 1) nis analysis followed the approved BE LBLOCA methodology for three and four loop plants (Reference 2). The analysis used ECOBRA/ TRAC MOD 7A, Revision 1, as documented in Reference 3.
Farley Units I and 2 are three loop plants similar in design to the VRA plant design used to demonstrate the application of the BE LBLOCA methodology in Reference 4. Unit I has an upflow barrel / baffle (B/B) vessel as shown in Figure 12 1 and Unit 2 has a downflow B/B vessel (Figure 12 2). The models are essendally identical except for the presence of gaps 14,15 and 16 which connect the downcomer with the B/B for Unit 2. Additionally, the B/B region represented by channel 9 is blocked at the top of Unit 2's section 3. The loop layout for both plants is shown in Figure 12 3. For the purpose of this uprate analysis, it was desirable to perform the full analysis for one unit and have it bound both units. An initial transient was performed for each unit, setting major plant conditions as shown in Table 121. Comparison studies were performed on each plant to determine the limiting configuration. These studies showed that Unit 2 was the limiting plant configuration.
With the limiting plant configuration, sensitivity studies were performed to verify the bounding ,
plant conditions. A Loss-Of-Offsite Power (LOOP) study confirmed the assumption that pumps offsite power available is more limiting. A SI injection study showed that the assumption of loss of a full train of SI was the limiting single failure assumption. These assumptions were used in the reference transient.
Initial calculations were performed using a CD=1.0 DECLG break. Further calculations showed inat a CD=1.0 Split break was the limiting reference transient. The results of these calculations are shown in Table 12-2. A plot of the PCT transient for the CD=1.0 Split case is shown in Figure 12-4. Split breaks have been determined to be more limiting in some cases because they result in a small downward core flow during blowdown (see following discussion).
12 1
NSD-SAE ESI-97-647
' Reference Transient Description The LOCA transient can be divided into time periods in which specific phenomena are occurring.
- A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the hot assembly undergoes. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below. Results are shown in Figures 12-4 to 1217.
Critical Heat Flux (CHF) Phase 4
Immediately following the cold leg rupture, the break flowrate is'subcoc'ed and high. The regions of the RCS with the hottest initial temperatures (core, upper plenum, upper head, and hot legs) begin to flash to steam within the first 0.5 seconds following the break. Flow in the core reverses, and the fuel rods begin to go through departure from nucleate boiling (DNB). Voiding in the core also causes the fission power to drop rapidly. The discharge flowrate decreases sharply as the break flow becomes two-phase (Figure 12 6). This phase is terminated when the water in the lower plenum and downcomer (DC) begin to fla)h.
Uoward Core Flow Phase Flashing in the lower plenum and pumped flow supplied by the intact loops re establishes upward core flow for a brief period of time (Figure 12 7). This phase ends as the lower plenum mass is depleted, the loops become two phase, and the intact loop pump head degrades because of two-phase conditions (Figure 12 8).
Downward Core Flow Phase Downward flow into the core begins as the pump head continues to be degraded and upward flow in the DC is firmly established (Figure 12-9).
Due to the downflow during this phase, tne cladding temperature was turned around at about 15 secunds after the initiation of the transient. For certain size split breaks, the fraction of break flow drawn from the core is smaller than the DECLG break, leading to poorer core cooling during this period. The accumulators on the intact loops begin to inject at 14 seconds after the break (Figure 12-10). Initially, the injected water is bypassed around the downcomer and out of the break. As the system pressure continues to' fall (Figure 12-11), the break flow and consequently the downward core flow are reduced. The vessel pressure reaches the containment
, pressure at the end of this phase, which occurs about 30 seconds after the initiation of the transient. The core begins to heat up as the system approaches containment pressure and the vessel begins to fill with ECCS water.
Refill Phase When the steam flow up the downcomer is sufficiently : iced, itc cold ECCS water begins to penetrate the downcomer (Figure 12-12) and refill the lower plenum (Figure 12-13). The refill 12 2
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A
NSD SAE ESI 97-647 design, which is covered by the SER.. The temperature ranges used for the accumulator and SI ECCS water were calculated using plant specific data collected for a year to determine the nominal and maximum values for ranging. All requirements outlined in the SER were confirmed for the Farley analysis. Among the required checks are the following:
- 2. Confirm normality of several key distributions used in the analysis.
The required calculation of core wide oxidation has been performed and the result is shown on Table 6.1.12 of Reference 1. The long term core cooling calculation did not use ECOBRAfrRAC and is discussed further in the response to RAI 18.
The time step assumptions used in the Parley analysis are listed in Table 12 3 and follow the approved methodology.
Summary The BE LBLOCA analysis for Farley Units 1 and 2 followed all of the guidelines and requirements as specified in the SER (Reference 2) and resulted in a peak cladding temperature
<2064'F, which meets the acceptance criteria. The operating ranges for major plant parameters are accounted for in the estimated PCT uncertainty. Table 12-2 shows some of these parameter ranges. A complete list is provided in the FSAR and in the response to RAI 13.
12-4 s
- NSD-SAE ESI 97-647
References:
- 1. WCAP-14723, "Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," January,1997,
- 2. Letter, R. C. Jones (USNRC) to N. J. Liparulo (E), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of Coolant Analysis," June 28,1996.
- 3. Letter, N. J. Liparuto (_W) to R. C. Jones (USNRC), " Revisions to Westinghouse Best.
Estimate Methodology," NTD-NRC 95 4575, October 13, 1995.
- 4. " Westinghouse Code Qualification Document for Best Estimate Loss of coolant Accident Analysis," WCAP-12945 P (Proprietary), Volumes I-V.
12-5
A NSD-SAE ESI-97-647 '
Lt Table 121 -
. Major Plant Parameter Initial Assumptions Used in the BE LBLOCA Analysis- !
for Farley Units I and 2 Power Uprate;
^
Parameter Initial Value Range Plant Initial Operating Conditions: I Reactor Power 100% of 5102% of
.2775 MWt -2775 MWt -
l Peak Linear Heat Rate (PLHR) Derived from desired Fo's 2.5 Tech Spec (TS)!!mit
- - Fn= 2.5 and maximum baseload Fo Hot Rod Avg Linear Heat Rate Derived from TS Fas 1.70 F ,= 1.7 Fluid Conditions Tavg 567.2*F 567.2 2 6'F s Tavg 5 577.2
- 6*F Pressurizer Pressure 2250 psia 2250
- 50 psi
- Loop Flow - 86,000 GPM 2: 86,000 GPM '
Accumulator Temperature 105'F 90s T.s 120*F
, Accumulator Pressure 640 psia 600 s P, s 680 psia Accumulator Volume 980 ft' 965 s V.5995 ft' s Accident Boundary Conditions j Offsite Power Availability On On or Off Single Failure Assumption Loss of one RHR -
pump Safety Injection Temperature Nominal (85'F) - 70 s SI Temp s 100 'F Safety Injection Delay . 12 sec. 512 seconds (no LOOP) s 27 sec. (LOOP)
Containment Pressure - - Bounded See Figure 12-17 i
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NSD-SAE-ESI 97-647 Table 12 2 Scoping Study Results for Farley Unit 2 Break Type Break Size' Renood 1 PCT ('F) Renood 2 PCT
('F)
DECLG 1,0 1810 1861 Split 0.8 1718 1650 Split 1.0 1871 1936 Split 1.2 1867 1778 Split 1.4 1718 1436
- Fraction of cold leg area (CD for DECLG)
Table 12 3 Time Step Assumptions for Farley Unit 2 Calculations
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NSD-SAE ESI-97-647
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NSD-SAE-ESI 97-647 1
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NSD-SAE ESI-97-647 30 L
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