ML20059F322
| ML20059F322 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 11/03/1993 |
| From: | Quirk J GENERAL ELECTRIC CO. |
| To: | Joshua Wilson Office of Nuclear Reactor Regulation |
| References | |
| MFN-180-93, NUDOCS 9311040238 | |
| Download: ML20059F322 (15) | |
Text
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e GENucIcar Energy 1
Genera!Bectnc Camnav 175 Car *ncr kenue. Sun Jcy, CA 95125 November 3, 1993 MFN 180-93 Docket No. STN 52-001 Document Control Desk U. S. Nuclear Regulatory Commission l Washington, D. C. 20555 !
Attention: Mr. Jerry N. Wilson, Acting Director ,,
Standardization Project Directorate !
Subject:
Submittal of Draft Revision of ABWR SSAR f Table 19.8-1 through 19.8-7
{
Centlemen:
Enclosed are thirty-four (34) copies of a draft revision of ABWR SSAR Tables 19.8-1 through 19.8-7 that have been annotated with cross-references to the inspections, tests, analyses and acceptance criteria (ITAAC). The ITAAC are !
contained in the GE Document 25AS447 " Certified Design Material for the ABWR."
The Section 19.8 tables list the ABWR design features identified by the plant probabilistic risk assessment (FRA) as making significant contributions to ,
plant risk reduction. The annotation for a particular design feature '
! indicates the ITAAC which will be used to confirm that the as-built facility
! has incorporated the particular design feature. !
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The attached material is draft; GE intends to include the final version in the ABWR SSAR amendment currently scheduled for submittal in mid-November 1993.
To support this schedule, GE would like to resolve any staff concerns on this annotated version of the table by November 10, 1993. NRC comments on a ;
compatible schedule will be appreciated.
As always, GE personnel will be available to provide clarification of this material and to answer any questions the staff may have. .
Sincerely, l .
l !
l.
l J. . u; rk, Project Manager A Certification ,
M'C 782, (408)925-6219 enclosures (34) cc: J. A. Beard-GE J. N. Fox-GE T. A. Boyce-NRC A. J . James-GE I J. D. Duncan-GE R. Louison-GE P. D. Fletcher-DOE F. M. Paradiso-GE >
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, 23A6100 Rev. 2 ABWR_ Stamiant SafetyAnalysis Report 19.8 Important Features identified by the ABWR PRA l Introduction The ABWR PRA has been resiewed to identify important design features,i.e., those features and actions that contribute significantly to the mitigation or prevention of a particular accident sequence or event scenario. These may be important contributions relating to 1
e System capability a Structures, systems, and components denoted by importance measures such as Fussell-Vesely ,
i i
a Bypass sequences (containment .>nd suppression pool) l a Features identified in SECY 93-087 s How the design meets containment performance goals a External events a Shutdown events i a important core damage sequences a What keeps core damage frequency (CDF) low a What has large uncertainty and in the extreme could become a significant contributor to CDF This section describes the logical process used to identify the important design features and provides the basis f or the importance of the feature. These design features are listed in Tables 19.8-1 through 19.8-7. These tables are annotated to show where the features 4 are addressed by ITAAC in the Certified Design Material (See Secdon 14.3). These ITAAC verify that the as-built facility correctly incorporates the annotated feature.
Logical Process Used to Select Irnportant Design Features ,
Although each design feature that can prevent or mitigate core damage is important to some degree and should be correctly and fully implemented, there are features that i
provide a greater degree of protection than others and can be considered more "important." For each initiating event (e.g., flood, fire, LOCA), there are components ;
or features that are more important than others for the prevention or mitigation of the event being evaluated. Where contributions to CDF have been determined by the calculation of Fussell-Vesely or Risk Achievement factors, these parameters can be used to identify the most important features. If the analysis does not result in the calculation ofimport:mce measures, other bases are used. For example, a single fea:ure that can important Features identified by the ABWR PRA - Draft - 19.8-1
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] , 23A6100 Rev. 3 ABWR Standard SafetyAnalysis Report l Table 19.8-1 Important Features from Level 1 Internal Events Analyses Feature Basis I
Capability to operate RCIC for eight hours This system with this capability provides the l
without AC power (2.4.4 2 hrs. only), and ability only means available to provide core cooling to override switchover to makeup water source with the reactor at high pressure and avoid core '
I from CST to suppression pool (2.4.4).This damage in the event of a station blackout.
l defines requirement for station battery capability to provide RCIC control power for l eight hours (2.4.4 2 hrs. only).
Combustion turbine generator connectable to Provides a diverse source of emergency AC ,
any of the three safety divisions and capable of power as added defense against loss of offsite ;
powering one complete set of normal safe power and diesel generator failure events. j shutdown loads (2.12.11). No plant support
, systems are needed to start or run the CTG (2.12.11). The CTG starts automatically and l safety-grade loads are to be added manually (2.12.1,2.12.11). ,
Operability of one high pressure core flooder Provides an independent and diverse means of (HPCF) loop independent of essential initiating emergency core cooling in the event of i l multiplexing system (2.' _). postulated common mode failures in the digital safety system logic and control (SSLC).
AC-independent Water Addition System, Provides an independent and diverse means of including a dedicated diesel (2.15.6) and achieving emergency core cooling in the event manually operable valves (2.15.6), to provide a of station power loss or failure of the diverse means of low pressure water injection engineered safety features to provide this into the reactor vessel. function. ,
Sufficient cooling capacity available in the RCW The redundant capability in each RCW/RSW system to provide seal and motor bearing division to successfully support ECCS functions :
cooling for ECCS core cooling pumps with one substantially lowers the calculated CDF. l RCW and one RSW system pump in each loop in each division and two RCW heat exchangers in ;
each division operating.
All piping systems, major systems components The designing of interfacing low pressure (pumps and valves), and subsystems connected systems to URS equal to RCPB pressure reduces ,
to the reactor coolant pressure boundary (RCPB) the possibility of an intersystem loss of coolant ,
which extend outside the primary containment accident and consequently the possibility of a boundary are designed to the extent practicable loss of coolant accident outside the '
to an ultimate rupture strength (URS) at least . containment.
equal to full RCPB pressure (2.4.1,2.4.2,2.4.4,
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2.2.2, 2.2.4, 2.6.1).
Redundant and diverse CRD scram capability The CRD scram system provides the first line of r consisting of both hydraulic and electric run-in defense against ATWS events. In addition, the ;
capabilities (2.2.2) with redundant and diverse redundancy and diversity incorporated in the ;
scram signals from the RPS (2.2.7) and ARIlogic CRD scram system significantly reduces the f (2.2.8). probability of an ATWS. l 19.8-28 Important features identified by the ABWR PRA - Draft !
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. 23A6100 Rev. 3 ABWR_ standard sarety Analysis neport Table 19.8-1 Important Features from Level 1 Internal l Events Analyses (Continued) l Feature Basis 'l Automatically initiated standby liquid control The automatic SLC and recirculation pump trip (SLC) system (2.2.4) and recirculation pump trip provides backup shutdown capability to the f (2.2.8) to provide backup shutdown capability in CRDs which substantially reduce the calculated [
event of failure to insert control rods. CDF associated with an ATWS event. l Three separated divisions of engineered safety The separated divisions of ESF provides three [
features, each containing both high and low complete divisions of redundant engineered pressure emergency core cooling systems as safety features which are the bases for the low ;
well as the capability to remove decav heat.The calculated CDF of the ABWR. l integrity of divisions is important. No high :
l pressure or high temperature piping lines i
should penetrate walls or floors separating two different safety divisions (Combination of: 2.4.1, :
2.4.2, 2.4.4,2.1.2,2.11.3,2.11.9,2.12.1,2.12.12, 2.12.13,2.12.14). Piping penetrations should be ,
qualified to the same differential pressure l
requirements as the walls or floors they
- l penetrate (2.15.10, 2.15.12).
Automatic Depressurization System to provide The ADS provides a reliable means of ;
access to low pressure core cooling injection depressurizir:g the reactor to permit core j systems (2.1.2). cooling with low pressure systems in the event high pressure systems fail. l Three emergency diesel generators, one The three emergency diesel generators provide i dedicated to each of the three safety divisions redundant sources of emergency AC power as I and each capable of powering the complete set added defense against loss of offsite power of normal safe shutdown loads in its division events.
t (2.12.13).
Four divisions of self-tested Safety System The four division SSLC provides reliable i Logic and Control instrumentation designed on defense against ATWS events as well as reliable
[ the basis of two out of four actuation logic (3.4). initiation of ESF core cooling and heat removal i systems. ;
Conduct of quarterly testing of the Essential This testing is conducted to discover faults that Multiplexing System and the Safety System are not identified by the continuous self-test ,
Logic and Control System. feature. The conduct of the quarterly testing [
substantially increases the reliability of the
! Essential Multiplexing System and the Safety i System Logic and Control System and the j subsequent contribution to the low calculated !
CDF.
Administrative actions to avoid common-cause Reduce the potential for common-cause failures failures noted in Subsection 19.9.8. to disable safety systems. :
l hrpc.' tant Features identified by the ABWR PRA - Draft 19.B 29 I
, . 23A6100 Rev. 3 'l ;
4 ABWR standardSafety Analysis Report i i
i Table 19.8-2 Important Features from Seismic Analyses ;
i Feature Basis t Seismic design of the containment (2.14.1) and Failure of seismic Category I structures could i lead directly to core damage because of j reactor building (2.15.10) and assurance that future modifications or additions to internal possible damage to ESF equipment. The j structures meet the requirements of Containment and the Reactor Building are the :
Subsection 3.8 if they are made in the vicinity of seismic Category I structures with the lowest i safety equipment. HCLPFs.
Seismic qualification of the station batteries, DC power is required for all safety-related f battery racks, and battery chargers (2.12.12). instrument and equipment control functions.
directly to core damage. :
Seismic qualification of the emergency AC in a severe seismic event, it is likely that offsite power system diesel generators (2.12.13),480V AC power will be lost and emergency AC power transformers (2.12.1), circuit breakers (2.12.1, will be the only source of AC power.The j 2.12.12), and motor control centers (2.12.1, components in the emergency AC power .
l 2.12.12). system with the lowest HCPLFs are the diesel I generators,480V transformers, circuit breakers, I
and motor control centers.
Seismic qualification of the plant service water in a severe seismic event, it is likely that offsite system service water pumps, room air AC power will be lost and emergency AC power conditioners, and pump house (2.11.3). will be the only source of AC power. The plant l l
service water system is required for diesel I generator cooling and othar cooling functions. l The components in the service water system r most sensitive to a seismic event are the service ,
water pumps, room air conditioners, and pump house. l Seismic qualification of SLC system boron in a severe seismic event, the ability to insert ;
solution tank and SLC pumps (2.2.4). control rods may be impaired due to seismic
[ ,
deformation of the fuel channels and the SLC ;
system may be the only means of reactivity ;
control. The most sensitive components in the ;
SLC system are the boron solution tank and the I
SLC pumps.
Seismic qualification of the ACIWA system ACIWA is seismic Category I and can provide. ,
including the pumps, valves, and water supply either vessel injection or drywell spray using ,
[2.15.6 (SSE only)].The collapse of the ACIWA equipment that does not require AC power. In l building (shed) should not prevent the pumps addition, support systems normally required for from starting and running [2.15.6 (SSE only)]. ECCS operation are not required for ACIWA
- l All needed valves for system operation can be operation. ACIWA is an important system in l
accessed and operated manually ( 2.15.6,2.4.1). preventing and mitigating severe accidents.
I 19.8-30 Important Features identified by the ABWR PRA - Draft i
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. 23A6100 Rev. 3 ABWR. standardSafety Analysis Report i
Table 19.8-2 Important Features from Seismic Analyses (Continued)
Feature Basis
, Seismic qualification of the RHR heat Seismic failure of RHR heat exchangers could i I
l exchangers (2.4.1). partially drain the suppression pool and flood {
the RHR rooms. RHR is needed for decay heat removal and water in the suppression pool a'
would provide fission product scrubbing in the
' event of core damage.
Seismic walkdown A seismic walkdown could identify seismic :
vulnerabilities which were not identified in the margins assessment.
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l Important Features identified by the ABWR PRA - Draft 19.8-31 I
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Table 19.8-3 Important Features from Fire Protection Analyses l 4
Feature Basis Fire detection and suppression systems are The use of these systems (not credited in the l provided throughout the plant (2.15.6). Fire analysis) will make core damage frequency
- detection methods include infrared and much less than the screening value of 1E-6. j product-of-combustion. Fire suppression .
systems include hand held fire ;
extinguishers, water hoses, foam and fire ,
sprinklers (2.15.6). FPS actuation is alarmed ,
in the control room (2.15.6). ;
The Remote Shutdown Panel with the ability The Remote Shutdown Panel provides an to control HPCFB, four SRVs, and two independent alternative means of achieving l divisions of RHR (2.2.6,2.1.2). safe shutdown of the reactor in the event that the control room becomes l s
uninhabitable due to a fire or other event.
The capability to operate the RCIC from The capability to operate a redundant and l outside the control room (2.4.4) and the diverse high pressure injection (RCIC) ,
capability to operate four SRVs from the system and the capability to operate a l l remote shutdown panel (2.2.6,2.1.2). redundant fourth SRV from outside the !
control room were required to meet the 1E-6 fire risk screening criterion.
l Design and maintenance of divisivital The integrity of the divisional fire barrier ;
separation by three hour rated fire barriers separation is required to meet the 1.0E-6 fire i of engineered safety features and their risk screening criterion. This assures that a l support systems including the intake fire in one division will not cause equipment :
structure (e. g., electrical power and cooling in another division to fail because of fire ;
water). Subsection 9A.5.5 under Special propagation between divisions. '
Cases-Fire Separation for Divisional ElectricalSystems lists the only areas of the plant where there is equipment from more than one safety division in a fire area. These l should be the only areas where multiple j divisions share the same fire area (2.15.10, i l
2.15.12).
Routing of piping or cable trays during the This design feature assures that the routing detailed design phase will conform with the of piping or cable trays will not invalidate fire area divisional assignment documented the requirement that all safety divisions are j in the fire hazard analysis (2.15.10,2.15.12). separated by three hour fire barriers.The integrity of the divisional fire barrier l separation is required to meet the 1E-6 fire risk screening criterion.
Design, maintenance and testing of smoke The prevention of the spread of smoke, hot l control systems (2.15.5c, 2.15.6). gasses, and fire suppressant from one fire division to another is implicit in the FIVE analysis as an important requirement to !
prevent adversely affecting safe shutdown -
capabilities, including operator actions. !
19.8-32 Important Features Identified by the ABWR PRA - Draft
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'3A6100 Rev. 3 ABWR_ standardsafety Analysis Report l~
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- Table 19.8-4 Important Features from Suppression Pool Bypass and Ex-Containment LOCA Analyses' ;
Feature Basis l i
l DW-WW vacuum breakers (2.14.1). Failure of a DW-WW vacuum breaker to close provides a significant bypass from the i drywellinto the wetwell airspace following a
$ drywell LOCA or if RPV failure occurs. This bypass pathway can release fission products
- directly to the atmosphere if high wetwell l pressure causes the containment rupture disk '
I to open.The consequence of a wcuum
] breaker failing to close and causing the !
rupture disk to open was <. <aluated in the !
PRA.
j Redundant Main Steam isolation Valves The MSL is very large compared to other l (MSIVs) (2.1.2). The MSIVs are pneumatic bypass pathways and a failure of both MSIVs operated, spring close, fail-closed designs in one steamline to close would provide a (2.1.2) actuated by redundant solenoids - large bypass pathway from the RPV to the through two-out-of-four logic (2.4.3). turbine building. Therefore, the failure of the i MSIVs to close would have a higher consequence from a given postulated event than other bypass pathways.
The SRV discharge lines are designed and A break in one of these lines in the wetwell fabricated to Quality Group C requirements . airspace could cause the containment l (2.1.2) and the welds in the wetwell region rupture disk to open and result in a pathway above the surface of the suppression pool are directly from the RPV to the environment.
non-destructively examined to the requirements of ASME Section Ill, Class 2 t
(2.1.2).
Normally closed sample lines and drywell if sample lines or purge lines are purge lines. inadvertently left open a bypass pathway can exist.
Redundant and seismically qualified CUW Minimize the potential for an ex-containment system isolation valves (2.6.1), qualified to LOCA to lead to core damage and potential l
close under postulated break offsite release. ,
l conditions 12.6.1,1.2(4)].
RPV drain line connection to CUW section improve the ability of the operator to line is 38.1 cm (15 in) or more above top of properly control RPV water level following 1 l active fuel (2.6.1). CUW suction line and unisolated CUW break.
connection to RPV is 1.52 m (5 ft.) or more l above top of active fuel (2.1.1).
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i important Features identified by the ASWR PRA
- ID!! 19.8-33 l
, 23A6100 Rev. 3 standardSafety Analysis Report ABWR.
i Table 19.8-4 Important Features from Suppression Pool Bypass and Ex-Containment LOCA Analyses (Continued)
Feature Basis Blowout panels in the RCIC and CUW Failure of the blowout panels during an ex- ,
divisional areas (2.15.10 As-built structural containment LOCA d Je to a break in a RCIC i evaluation). or CUW line could result in the pressurization of these divisional areas that could impact ,
equipment in adjacent ar:,as and result in a second electrical division being unavailable. r i
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l 19.8-34 Important Tectures identified by the ABWR PRA - Draft w
23A6100 Rev. 3 ABWR Standard Safety Analysis Report Table 19.8-5 Important Features from Flooding Andyses j i
Feature Basis Equipment for each safety division is located Assuming a flood has occurred and other !
within enmpartmerits designed to prevent water mitigation features have failed, this single from a flood from propagating from one division design feature prevents flooding in one l
to another (2.15.10,2.15.12).This includes features division from affecting another division.
such as watertight doors and sealed cable l penetrations (2.15.10, 2.15.12).
4 Floor drains in all upper floors of reactor and Assuming a flood has occurred and other ,
q l control buildings (2.9.1). mitigation features have failed, this single ,
4 feature assures that flood waters on upper l j floors of the reactor and control buildings will ;
! flow to lower floors thereby preventing the i
! failure of important equipment on that floor ,
j and allow other features on lower floors to j mitigate the flood (e.g., sump pumps, j watertight doors). !
l Water level sensors in RCW/RSW rooms (2.15.12) Assuming a flood has occurred, the water ,
i and logic in the control building to alert operator level sensors and logic are the only automatic ;
and trip RSW pumps (2.11.9 Interface Req.)and features that can identify and terminate l 1 close valves in affected RSW division (2.11.9). flooding in the RCW rooms.
4 The reactor building corridor on floor B3F is large Assuming a flood has occurred and other
! enough to contain the largest flood sources in the mitigation features have failed, this feature reactor building (condensate storage tank or prevents any flood in the reactor building that j suppression pool) (2.15.10). flows to the corridor from affecting any safe '
shutdown equipment in the reactor building by isolating the water in the 83F corridor. ;
Anti-siphon capability in RSW systems (2.11.9 Anti-siphon capability will prevent a control i building flood from continuing to siphon interface Req.).
water after the pumps have been stopped. l Failure of this capability could increase the I chances of some floods leading to core damage. ;
Reactor Building sumps on floor B1F have overfill Assuming the failure of the sump pumps or a lines to the B3F corridor. Loop seals are provided flood that exceeds the capacity of the sump on the overfilllines. pumps, these overfill lines prevent flood j water in one division from propagating to I another division. Loop seals are provided to j
, preserve the integrity of the secondary l i containment.
, Suildings other than the Turbine, Control, and The screening analysis indicated that the Reactor Building do not contain equipment that flooding enalysis only needed to address can be used to achieve safe shutdown and internal flooding from sources in the Turbine, flooding in those buildings cannot propagate to Control, and Reactor Buildings.lf this is not buildings which contain safe shutdown equipment the case, the basic flooding analysis could be (Multiple ITAAC entries define ABWR design). invalidated.
Important Features identified by the ABWR PRA - Draft 19.8-35
. 23A6100 Rev. 3
} ABWR, Studard Safety Analysis Repcit l
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Table 19.8-5 Important Features from Flooding Analyses (Continued) i Feature Basis ,
, Operator check on each shift that watertight doors A watertight door must be dogged to assure ,
d are closed and dogged. that it will provide full protection in the event !
i of a flood.
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High pressure or high temperature line[ not This single design feature prevents the failure
, routed through floors or walls separatili; two of one division of a system ultimately .
different safety divisions (2.15.10 Divisii.nal resulting in the flooding and disabling of a !
Separation). second division. l 1
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- 23A6100 Rev. 3 ABWR Standard Safety Analysis Report Table 19.8-6 Important Features From Shutdown Events Analyses Feature Basis Decay Heat Removal Shutdown cooling (SDC) mode of the RHR RHR(SDC)is capable of both removing decay system (2.4.1). heat and ensuring that the core is covered with l
water. SDC is the normally used and preferred method of decay heat removal (DHR) during shutdown.
j Reactor service water (RSW) system (2.11.9). Failure of the RSW system would disable the principal RHR system. The RSW removes heat from the RHR and other systems and transfers it to the ultimate heat sink.
Ultimate heat sink (UHS) (4.1). The UHS rejects decay heat to the environment l
from the RHR/RCW/RSW systems.
Inventory Control The low pressure core flood mode of the RHR The low pressure core flood mode of the RHR l system (2.4.1). system can supply makeup to the reactor with the reactor at low pressure.
The CRD system pumps which can supply water The CRD system pumps are capable of l to the core through the CRD purge flow (2 2.2). providing makeup to the reactor at high and low pressures to ensure the coce is covered.
l High pressure core flooders (HPCF) (2.4.2). The HPCF is capable of providing makeup to the reactor at high and low pressure to ensure the core remains covered.
AC-independent Water Addition (ACIWA) The ACIWA can supply makeup to the reactor l System (2.15.6,2.4.1). with the reactor at low pressure.
l RPV isolation on low water level (2.4.3). The isolation of lines connected to the RPV on a low water level signal prevent uncovering the fuel for many potential RPV drain down events.
Permissives and inhibits associated with the The permissives and inhibits associated with the RHR Mode Switch (2.7.1). RHR Mode switch ensures that valve line ups l
are correct for most RHR functions thereby preventing inadvertent diversion of water from the RPV.
RHR Valve Interlocks (2.4.1). The RHR valve interlocks prevent low pressure l
HHR piping connected to high pressure systems from being exposed to high pressures.
l RPV Level indication (2.1.2). The RPV level instrumentation informs the operator of RPV level and allows automatic initiation of ECCS pumpc and closure of RPV isolation valves on low water level.
Important Features identified by the ABWR PRA - Draft 19.8-3 7
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. 23A6100 Rev. 3 .
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Table 19.8-6 Important Features From Shutdown Events Analyses (Continued) l Feature Basis l l RIP Diffuser Plug (2.1.3) RIP maintenance during shutdown requires a l l temporary plug be installed in the RIP diffuser when RIP impeller, shaft and motor are i removed. The plug is designed so it can not be removed unless the RIP motor housing bottom [
l cover is in place.
1 b Reactivity Control !
l RPS High Flux Trip (Set Down) (2.2.7). The RPS high flux trip automatically inserts withdrawn CRDs at a specified flux level to prevent criticality. l l CRD Brake (2.2.2). Tim brake system on the CRDs prevents ejection f l
of a CRD which could cause criticality.
l Refueling Interlocks (2.2.1,2.5.5). When the reactor Mode switch is placed in the f REFUEL position, no fuel assembly can be :
hoisted over the RPV if a CRD blade has been f removed. ,
Containment Integrity l Automatic isolation of secondary containment The automatic isolation of the secondary .;
l l (Modes 3 and 4) (2.4.3). containment on a specified high radiation signal prevents release of radioactivity to the environs. i l SGTS (2.14.4). The SGTS processes gasses before release to the atmosphere, j Electrical Power f Three physically and electrically independent The three divisions of safety-related electric ;
l divisions of safety-related power (2.12.1). power allows for one division to be in i maintenance and still mitigate a single active ;
failure in another division. (
Four onsite sources of AC power (three EDGs The four sources of onsite AC power backs up l and one CTG) (2.12.13,2.12.11). offsite power and ensures power will be ,
available to safe shutdown equipment. I 1
Redundant offsite power sources allow for the Two independent offsite sources of AC power (2.12.1 Interface Req.). loss of one offsite power source without losing power for decay heat removal during shutdown.
19.8-38 Important Features identified by the ABWR PRA - Draft
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. 23AG100 Rev. 3 ABWR standardSafetyAnalysis Report 1 i J
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. Table 19.8-7 Key Severe Accident Parameters '
i Parameter Description Value Relates to What Feature?
Core Power 3926 MW Containment Performance _ l; El. of Top of Fuel 9.05 m Containment Performance Normal Water Level 13.26 m Containment Performance ADS Area 0.07m2 Vessel Depressurization !
l Containment Leak Rate 0.5% per day (2.14.1) Containment Performance Containment Service Level C 0.77 MPa Containment Performance f Containment Ult. Strength 1.025 MPa Containment Performance l; Total Zr in Core 72,550 kg Containment Performanc, Sacrificial Concrete ;
Calcium Carbonate Content <4 weight percent Basaltic Concrete ;
I (2.14.1) l Height of Layer 1.5 m (2.14.1) Sacrificial Concrete l l Pedestal Concrete Thickness 1.64 m (2.14.1) Pedestal !
Compartment Volume i
Lower Drywell 1860 m3 Containment Performance j Upper Drywell 5490 m3 Containment Performance j Wetwell 585 m3 Containment Performance {
t Floor Area ,
Lower Drywell 88 m2 Lower Drywell j Tolerance of Vacuum Breaker Po- 0.9cm Vacuum Breaker j sition Switch ;
i Overflow Elevation ;
I LDW to Wetwell -4.55 m Lower Drywell i f
UDW to Wetwell 7.35 m Lower Drywell LDW to UDW vent area 11.3 m2 (2.14.1)' Connecting Vents ,
l Lower Drywell Flooder .
Elevation -10.5 m Lower Drywell Flooder .
Area per valve .0081 m2 Lower Drywell Flooder Plug Temperature 533 K Lower Drywell Flooder Important Features identified by the ABWR PRA - Draft 19.8-39/40 I t
, 23A3100 Rev. 3 ABWR standantSafetyAnalysis Report O
l Table 19.8-7 Key Severe Accident Parameters (Continued) )
Parameter Description Value Relates to What Feature?
I Suppression Pool Mass 3.6 x 106 kg Containment Performance COPS Equivalent Flow Area of Disk 0.2 m (8") COPE ,
Diameter of Piping 0.25 m (10") COPS Setpoint J.72 MPa (90 psig) COPS (2.14.6)
Tolerance at nom temp 5% COPS Effect of temp on setpoint 2% per 100 F COPS ;
Firewater Addition System j i
Injection Locations Vessel and Drywell ACIWA i (2.15.6, 2.4.1) l Runout flow 0.06 m3/sec ACIWA Flow rate at 90 psig 0.04 m3/sec ACIWA ;
Corium shield ;
t I l Height 0.4 m (2.14.1) Corium shield l Depth 0.4 m (2.14.1) Corium shield [
t l Material Alumina (2.14.1) Corium shield ,
l Channel Length for Equip- 1 m (2.14.1) Corium shield I ment Sump Shield .
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l 19.640 Important Features identified by the ABWR PRA - Draft l
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