ML20096F024

Forwards Info Re Advanced BWR Containment Steam Bypass Leakage Capability.Sensitivity Study Results Demonstrate That Currently Specified Bypass Leakage Capability of 0.05 ft2 Is Not at High Point of Cliff

ML20096F024
Person / Time
Site:	05200001
Issue date:	05/01/1992
From:	Fox J GENERAL ELECTRIC CO.
To:	William Burton, Kudrick J, Poslusny C NRC
References
NUDOCS 9205200150
Download: ML20096F024 (11)

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ABWR CONTAINMENT STEAM BYPASS CAPABILITY BACKOROUND currently, a steam bypass leakage capability (A/K(1/2)) of 0.05 ft 2 '

is specified for the ABWR design. And consistent with the SRP requirements, the ABWR Technical Specifications will define-and require that maxinum leakage during periodic leakago rate tests shall

- be less than 10% of the design maximum bypass leakage capability.

During the October '91 and December '91 meetings with the staff in San Jose, GE described and explained the basis for specifying an allowable steam bypass leakage capability of 0.05 ft 2 for the ABWR design. Tha staff understood and recognised the basis of the currently defined bypass capacity of 0.05 ft 2. However, the staff requested QF,to confirm that 0.05 ft3 is not at the high point of cliff, considering a full spectrum of primary system break sizes.

GE, in response to this staff's request, undertook a task to perform sensitivity study evaluating steam bypass leakage capability over a full rpectrum of pipe breaks and confirm that 0.05 ft2 is not the high point of cliff. Additional objectives of this sensitivity study were to assesa feasibility of achieving leakage capability greater than 0.05 ft2 at both the containment design pressure as well as the overpressure protection set point values.

The following paragraphs describe and discuss results from this sensitivity stady.

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t m 01 '92 12: 33Pt1 P.14 STEAM BYFAB8 LEAKAGE If a direct leakage path were to exist between the drywell and the vetwell airspace Juring a loss-of-coolant accident (LOCA) event, the laaking steam bypasning the watwell suppression pool would produco rapid pressurization of the wetvell airspace. To mitigate the consequences of any steam which bypasses the suppression pool, the ABwx design provides safety grade drywell and wetwell spray systems.

Emergency Procedure Guidelines (EFGs) defining operator actions for controlling containment pressure, as necessary, have specified for the ABWR design.

For a given primary system break area, the maximum allowable leakage capacity can be determined when the containment pressure reaches the design pressure at the end of reactor blowdown. The most limiting conditions would occur for those primary system break sizes which do not cause rapid reacter depressurization enabling low pressure ECCS systems to start providing reactor vessel inventory makeup.

1. ABWR DEEIGN FEATURRS A. Eccs cor.ficurat),gn The ABWR ECCO design configuration, as a minimum, comprises of the following 1

Reactor Core Isolation Cooling (RCIC) Loop 2

High Pressure Core Flooder (HPCF) Loops 3

Low (RMR)Pressure Loops. Flooder (LPFL) mode of Residual Heat Removal C Two loops (RHR(B) and RHR(C)) have provision to operate in drywell/wetwell spray mode to remove heat from the

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containment.

1 Automatic Depressurization system (ADS)

(Independent of any other ECes)

The Ecca is separated into throe independent functional divisions as follows Division At RCIC + 1 RHR(LPFL) (RHR injects into FWL)

Division B: 1 HPCF + 1 RHR(LPFL/ Spray (Separato RHR injection nozzle)

Division C: 1 HPCF + 1 RHR(LPFL/ Spray) (Separate RHR injection nozzle)

In the event of a break in a pipe that is not part of the ECCS and allowing for single active component failure, tho available combination of ECCS equipaont shall be as followt (a) One HPCF + RCIC + two LPFL + all ADS valveJr or (b) Two HPCF + three LPFL + all ADS valvest or (c) Two HPCF + RCIC + three LPFL + all ADS valves minus one In the event of a break in a pipe that is a part of ECCS, and allowing for a single active component failure, the combination of Ecc8 equipment availablo shall be the ECCS equipment listed aYove minus the ECCS in which the break is assumed.

B. Eninganev Proemdure Guidelines tmergency Procedure Guidelines (EPGs) defining operator actions to control containment pressuro under'LOCA conditions have besn specified for the ABWR design in Chapter 18 of the DSAR. These guidelines specify operator actions and the conditions undLr which those actions can be undertaken, if that becomes necessary to control containment pressure under LCCA conditions. These actions, which are primarily I

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'yY0%'92 12134Pt1 P.16 symptom oriented, include actuation of drywell/watwell sprays and emergency depressuritation of the reactor pressure vessel, as shown in Figure 1. The ABWR design has provision for initiating drywell/watwell sprays independent of RHR system by using Firewater Addition system.

C. Drvwe11/Watwa11 sorgyL In the current design, each RHR pump has a rated flow capacity of 4200 gpm. When operating in spray mode the pump flow is split between the wetwell and drywell sprays. The ABWR design limits vetwell spray sparger flow to 500 gpm and drywell sparger flow to 3700 gpm. The pump maximum runout flow is 5000 gpm.

2, ANALY5F,8 Engineering analyses to evaluate steam bypass capability of the ABWR deJign were performed using approved engineering computer program, steam bypass capability at both the containment design p& assure value as well as the overpressure protection set point value

- was evaluated.

A full spectrum of primary system break sizes (0.01 through 1.0 f t ) was considered and evaluated, including both steam and liquid 2

breaks. It is to be noted that 1.0 ft2 is the largest primary system break area for the ABWR design. Operator actions, as permissible by

.the EPGs, were factored into these analyses. Thase actions included actuation of watwell sprays (when wetwell airspace pressure reaches 0.728 kg/cm2 g or 25 psia) and emergency depressurization of the reactor pressure vessel (when watwell airspace pressure reaches 1.54.kg/cm2 g or 36.6 psia), see Figure 1. For each case analyzed, available EOCS combination was determined considoring a worst single active component failure.

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IpY01'92 12134Ft1 P.17 Heat loss at the upper drywell and wetwell airspace concrete walls was modeled in the an' alysis. Heat loss at the lower drywell walls and suppression pool boundary walls was neglected.

A minimum time delay of 800 seconda for actuation of votwell spray was assumed. Drywell sprayo were not considered and modeled, because the calculated LOCA conditions in the drywell happened to be outside the EPGs specified drywell spray initiation range.

3. RESULTS 3.1 Centainment Desien Bypass C&2 ability Analyses were performed to determine maximum allowable leakage capacity of the ABWR design. For a given primary syst?m break area, the maximum allowable leakage capacity is determined when the containment pressure remains below the design pressure (60 paia) at the end of blowdown. These analyses considered and evaluated breaks in Feedwater line, in Main steam line, in LPFL injection line, and in HPCF injection line.

Results from these analyses which were performed te determine the allowable leakage between the drywell and the wetwell airspace during a primary system pipe break are shown in Figure 2. These resulta show the alloweble leakage capacity (A/K(1/2)) as a function of primary system break area. A is the actual area of the leakage flow path and K is the total geometric loss coefficient associated with the leakage flow path.

As seen from Figure ..., the maximum allowable leakage capacity is at A/K(3/2) = 0.1 ft2 . Since a typical geometric loss factor would be 3 or greater, the maximum allowable leakage flow path area would be

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MM 01 '92 12 ?5M1 P.18 0.173 ft 2, 3.2 Containgent_hygass Canability At Rueture Disc Set Point The maximum allowable leakage capacity is determined when for a given primary system break area the containment pressure remains below the rupture disc set point pressure (105 psia). Both liquid and steam breaks were considered and analyzed. A maximum allowable leakage capacity at A/K(1/2) = 0.35 ft2 was determined. Taking a value of 3 or greater for K, the maximum ellowable leakage path area would be 0.6 tt 2.

STEAM BYPASS CAPABILITY INCREASB varicas alternatives which would assist in increasing the current bypass capability limit were assessed for their merit and demerit. The alternatives which were assessed are AlternatiyE_A1 Vessel Cooldown Rate Greater Than 1000F/h Alternative Bt Single vs Two Valves in Series per Penetration Alternative C Increasing DW/WW Spray Capacities Alternative Dr Less Restrictive DW Spray initiation range Alternative A was considered and factored into the sensitivity study described above. Vessel cooldown rates ranging from 1000F/h to full ADS actuation were modeled and analyzed.

Alternative B was assessed for its merit and denerit for increasing the bypass capability. 7t is believed that two (simple check) valves in series are not expected to provide any significant improvement in the leakage area when compared to that with a single (simple check)

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. valve. On the other hand, two valves in series will make valve inspection and maintenance routine more time censuming which would have significant impact on plant outage schedule and plant maintenance cost.

Alternative c which would result in more steam quenching capacity, obviously, will be helpful in improving the ABWR steam bypass capability. It was determined that a sizeable upgrading of the current RHR design hardware will be necessary, in order for achieving any substantial increase in the DW/WW spray capacities.

Alternative D was judged to have a strong potential for providing a substantial increase in the steam bypass leakage capability with very minimum, or no impact, on the current ABWR RHR hardware design. The DW spray initiation region currently defined in the EPas is based on somewhat overly conservative initial conditions and assumptions. It appeared feasible to make the currant DW spray initiation region less L

restrictivo, oy reviewing and eliminating eny undue conservatism.

This alternative could be further evaluated for its effectiveness in increasing the bypass leakage capability, if that becomes desirable.

CONCLUS%0N The sensitivity study results presented and discussed above demonstrate that the currently specified bypass leakage capability of 0.05 ft2 is not at the high point of cliff, and there is substantial margin f. the ABWR design steam bypass capsbility.

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