ML20056F283
| ML20056F283 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/23/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9308260296 | |
| Download: ML20056F283 (32) | |
Text
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Gene:at[Ieanc Company 175 Cunv Annue, San Jow, CA 95125 August 23,1993 Docket No. STN 52-001 I
Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - Resolution of Severe Accident Issues *
Dear Chet:
Enclosed is a SSAR markup which provides resolution of several severe accident issues that have been raised by the staff. This markup will be included in the next amendment for the corresponding chapters.
Please provide copies of this transmittal to Bob Palla and John Monninger.
Sincerely, 3af ack Fox Advanced Reactor Programs cc: Alan Beard (GE)
Carol Buchholz (GE)
Jack Duncan (GE)
Norman Fletcher (DOE) unw 9308260296 930823 3 9 '
PDR ADOCK 05200001 0 / I A PDR $
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,. i Item 6C: Design information for ACIWA This material has been provided a a mark-up for 5.4.7.1.1.10. The system flow rate is given in 5.4.7.1.1.10.3.
Item D: ITAAC for Vacuum Breakers The ITAAC material for the vacuum breakers identifying the indication and alarm of the vacuum breaker position has been included in Tier 1 Section 2.14.1. This material was transmitted to the staff on 8/18/93 by Carol Buchholz. The vacuum breakers are included on the second page of the material for section 2.14.1.
Item E: Testing of Lower Dgwell Flooder There is a discrepancy between the testing interval specified for the lower drywell ficoder in subsections 9.5.12.4 and 19K.11.4. The correct interval is 2 outages as specified in 19K.11.4. The correction has been made in 9.5.12.4 as shown.
Item F: Purpose of stainless steel disk in Lower Drywell Flooder GE was asked to clarify the purpose of the stainless steel disk in the lower drywell flooder. The disk is adjacent to the water from the suppression pool. The purpose of this disk is to prevent corrosion of the Teflon or fusible material in the lower drywell flooder. This has been indicated in the design description for the lower drywell flooder in s ibsection 9.5.12.2.
Item H: Design Description Material GE has been requested to review the severe accident analysis for important system information that should be added to the system design descriptions in the SSAR to ensure that the system engineers and COL applicants have the benefit of any insights from the severe accident studies. To do this, the severe accidents insights were examined to determine those features which were determined to be important.
Then, the appropriate information was extracted from Appendix 19E (the severe accident analysis). The outcome of this process is given in the form of mark-up material to the SSAR. This mark-ups are provided as an attachment to this memo.
Note that the description for the COPS system was provided previously as a pilot package and is not included here. The COPS description has been incorporated in i Amendment 32 of the SSAR.
CEB93-26-2
23A6100 Rsv.1 ABWR standard sorary Anatysis a, port The ADS utilizes selected SRVs for depressurization of the reactor as described in Section 6.3. Each of the SRVs utilized for automatic depressurization is equipped with an air accumulator and check valve arrangement. These accumulators assure that the valves can be held open following failure of the air supply rc, the accumulators. The :
accumulator capacity is sufficient for one actuation at &ywell design pressure or five actuations at normal dnwell pressure.
Each SRV discharges steam through a discharge line to a point below minimum water level in the suppression pool. The SRV discharge lines are classified as Quality Group C and Seismic Category I. The SRV discharge lines in the wetwell air space are classified as Quality Group C and Seismic Catgory I, all welds shall be nondestructh ely examined to the requirements for ASME Boiler and Pressure Vessel Code,Section III, Class 2 piping. SRV discharge piping from the SRV to the suppression pool consists of two pans.
The first is attached at one end to the SRV and at its other end to the diaphragm floor penetration, which acts as a pipe anchor. The second part of the SRV discharge piping extends from the diaphragm floor penetration to the SRV quencherin the suppression pool. Because the diaphragm floor acts as an anchor arr this part of the line, it is physically decoupled from the main steam header.
As a part of the preoperational and startup testing of the main steamlines, movement of the SRV discharge lines will be monitored.
The SRV discharge piping is designed to limit valve outlet pressure to approximately 40% of maximum valve inlet pressure with the valve wide open. Water in the line more than a few feet above suppression pool water level would cause excessive pressure at the valve discharge when the valve is agam opened. For this reason, two vacuum reliefvalves are provided on each SRV discharge line to prevent drawing an excesshe amount of water into the line as a result of steam condensation following termination of relief operation. The SRVs are located on the main steamline piping rather than on the reactor vessel top head, primarily to simplify the discharge piping to the pool and to avoid the necessity of having to remove sections of this piping when the reactor head is removed for refueling. In addition, valves located on the steamlines are more accessible during a shutdown for valve maintenance.
The ADS automatically depressurizes the nuclear system sufficiently to permit the LPFL c mode of the RHR System to operate as a backup for the HPCF. Further descriptions of 9 ( -
t E # pried he operation of the automatic depressunzation feature are presented in Section and Subsection 7.3.1.
c(E[ y N 5.2.2.4.2 Design Parameters The specified operating transients for components within the RCPB are presented in Subsection 3.9.1. Subsection 3.7.1 provides a discussion of the input criteria for design of Seismic Category I structures, systems, and components. The design requirements 5.26 Jntegnty of Reactor Coolant Preesure Boundery- Amerutment 31
Insert Vessel Depressurization:
In addition to playing a major role in preventing core damage, depressurization of the RPV (either manually, automatically, or as a result of a LOCA) can help mitigate the consequences of severe accidents in which fuel melting and vessel failure occur. If the RPV were to fail at an elevated pressure (greater than approximately 14 kg/cm2g) high pressure melt ejection could occur resulting in fragmented core debris being transported into the upper drywell. The resulting heatup of the upper i drywell could pressurize and fail the drywell. This failure mechanism is eliminated if the RPV is depressurized. The opening of a single SRV is capable of depressurizing the vessel sufHeiently to prevent high pressure ,
melt ejection.
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5.4.7.1.1.10 AC-Independent Water Addition [jg g gg,4 The AC-independent water addition mode (Alternating Current independent) of the ,
(( . RHR System provides a means for introducing water from the Fire Protection System [g,
/ 6 i MPS) directly into the reactor pressure vessellto the dr)well spray header 6egraded i C/ plant conditions when AC power is not available from either onsite or offsite sources. ,
I The RHR System provides the piping and valves which connect the FPS piping with the RHR loop C pump discharge piping. The manual valves in this line permit adding water from the FPS to the RHR System if the RHR System is not operable. The primary means for supplying water through this connection is by use of the diesel.drhen pump in the 7 ? FPS. A backup to this pump is provided by alconnection on the outside of the reactor 7
f j-c @] building?which allows hookup of the FPS to a Ere truck pump.
Tbgessel injection mode is intended to prevent core aimage during on blackout 45' afiter thNSCIC System has stopd operating,'and to prmide an In - I core melt 8 preventioh amsm during a severe accid t conditio AGindependent water addition m is not Mtuated in tim prevpecore daarage, core m gand vesselfail , then it eI r drywell en initiated adds r{the corium in w4ter to containment, thheby41owing th are rise.
! \
The dryw spraymode / en)s<high temperatures in upper drywiljand adds additional ter to the ntamment,which cases the covamment therrAal mass hd slows the pressunzadon rate. Add [tionally,ellsk the provides
. fidion oductsfrubbing to reduce fission product release in the toffailureof thr.drywell cad.
Operation of the AC-independent water addition mode is entirely manual. All of the valves which must be opened or closed during fire water addition are located within the same ECCS salve room. The connection to add water using a fire truck pump is located pf outside the reactor building at grade level 1/
5.4.7.1.2 Design Bwis for isodation of RHR System from Reactor Coolant System The Ion pressure portions of the RHR System are isolated from full reactor pressure whenever the primary system pressure is above the RHR System design pressure (see Subsection 5.4.7.1.3 for details). In addition, automatic isolation occurs for reasons of mamtammg water inventory which are unrelated to line pressure raung. Alow water level signal closes the RHR contamment isolation valves that are provided for the shutdown cooling suction. Subsection 5.2.5 provides an explanation of the Leak Detection System and the isolation signal [see Subsection 5.2.5.2.1 (12) and Table 5.4-6].
The RHR pumps are protected agamst damage from a closed discharge valve 'tw means of automatic muumum flow vahes which open on low m=hhe flow and close on high m2mbne flow.
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8 76 N 1 4/ Y Insert:
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Figure 5.4-10 shows the connections from either the diesel-driven pumps or the fire truck to the RHR system. The connections to the diesel-driven pump are in the RHR valve room. Opening valves F101 and F102 allows water to flow from the FPS into the RHR piping. Periodic stroke testing of these valves is required by Table 3.9-8 to ensure valve operability. The fire truck connection is located outside the reactor building at grade level.
Both connections to the RHR system are protected by a check valve (F100 and F104 for the diesel-driven pump and the fire truck, respectively) to insure that RCS pressurization does not result in a breach of the injection path.
5.4.7.LL10.1 Vessel Igjection Mode of ACIWA The pnmary injection path for the ACIWA system is into the vessel via the LPFL header. For injection to occur, the RPV must be at low pressure. The purpose of vessel injection is to prevent core damage or, if core damage has already occurred, to terminate melt progression. Melt progression can potentially be terminated in-vessel if the debris has not failed the bottom of the vessel. After vessel failure, initiation of the vessel injection mode of the ACIWA system will cover the debris in the lower drywell with water.
If the vesselinjection mode of the ACIWA system is not initiated in time to prevent core damage,its use can mitigate the consequences of core damage by enhancing cooling, preventing radiative heating from the debris and adding thermal mass to the containment. Ifinjection is initiated prior to vessel failure, melt progression can be arrested in-vessel. However, if vessel failure occurs, debris will relocate from the vessel. If vessel failure occurs at low pressure (less than approximately 14 kg/cm2g), the debris will relocate only into the lower drywell..After vessel failure, water injected into the vessel will flow cut of the vessel breach into the lower drywell. Water flowing into the lower drywell will cover the core debris and enhance debris cooling.
Injection by the ACIWA system is terminated during a severe accident when the water levelin the containment reaches the bottom of the vessel.
Higher water levels could lead to a situation in which the piping of the Containment Overpressure Protection System (COPS) could be jeopardized. COPS activation is expected in core damage scenarios in which containment heat removal is lost and not recovered. If the suppression pool water level is near the COPS elevation when the rupture disk opens, water could potentially enter the COPS piping and impart signiScant water hnmmer loads. These loads are precluded by terminating water addition when the containment water level reaches the bottom of the RPV which is a few meters below the rupture disk. Another reason for termmating injection by the ACIWA system is the reduction in free space available in the wetwell for non-condensables as the suppression pool level rises. Reducing the non-condenaable volume increases contmhment
y}
pressure. Terminating injection at the bottom of the RPV approximately balances the pressure reduction due to heat absorption by the sprays and pressure increase due to non-condensable compression in the wetwell.
If vessel failure occurs with the RPV at an elevated pressure, high pressure melt ejection could occur resulting in fragmented core debris being transported into the upper drywell. Water injected into the vessel by the ACIWA system cannot reach this debris. In this scenario the drywell i spray mode of the ACIWA system must be used. The drywell spray mode is described in Subsection 5.4.7.1.1.10.2.
5.4.7.1.1.10.2 Drywell Spray Mode of ACIWA :
i The alternate injection path for the ACIWA system is into the drywell spray header. The conditions in which drywell spray mode is used are described in the Emergency Procedure Guidelines in Appendix 18A. The purpose of drywell spray injection is to mitigate the consequences of core l dnmage and to supply water to ex-vessel debris. l l
The water sprayed into the upper drywell absorbs heat from the RPV outer surfaces and the debris which relocates into the upper drywell, if any, upon vessel failure at high pressure. Cooling of the upper drywell prevents l overtemperature failure of the seals. Water which collects on the upper ;
drywell floor is directed into the wetwell through the connecting vents.
The suppression pool water level will eventually rise to the point of overflowing into the pedestal region. When overflow occurs, the debris in the lower drywell will be covered with water.
Drywell spray operation provides significant mitigation of suppression pool bypass events in which the bypass path includes the drywell. The incoming water absorbs heat and condense steam. While the heat absorption is not as efficient nor as extensive as what would occur if the suppression pool was not bypassed, the time to COPS activation.or containment failure can be delayed significantly. This delay results in a significant reduction in the radioactive release due to fission product decay and natural removal mechanisms.
The water sprayed into the upper drywell also scrubs fission products which are in the drywell airspace. Scrubbing reduces the amount of radioactive material which is available for release from the containment.
Drywell spray injection is terminated when the containment water level reaches the bottom of the vessel. The basis for termination is the same as that for the vesselinjection mode of the ACIWA system as described in Subsection 5.4.7.1.1.10.1
g 5.4.7.1.1.10.3 ACIWA Flow Rate The water flow rate of the ACIWA system has been selected to optimize the containment pressure response and slow the rate of containment ,
pressurization after the onset of core damage. The flow rate supplied to the i RHR System by either the diesel-driven pump or the fire pump truck is !
between 0.04 m3/sec (630 gpm) and 0.06 m3/sec (950 gpai) for conditions between runout and a back precsure equal to the COPS setpoint. This flow rate is sufficient to absorb decay heat while maximizing the time until the water level reaches the bottem of the vessel, at which point water addition is ternanated.
Flow rates outside the specified range will decrease the time to COPS actuation in situations in which containment heat removal is not recovered. Lower dow rates will result in some of the incoming water being vaporized, thereby increasing the rate of contaimnent pressurization.
Higher flow rates will decrease the length of time until the water level reaches the bottom of the RPV and flow is terminated. Containment pressurization ensues shortly after flow termination as the noncondenaables are purged into the wetwell and net steam production begins. Therefore, the optimni injection flow rate is the amount that can just absorb the generated heat without exceeding saturated liquid conditions at containment pressure.
5.4.7.1.1.10.4 Containment Performance Without ACIWA The ACIWA mode of the RHR System pmvides manual capability to prevent core damage when all emergency core cooling systems are lost. If core damage occurs and heat removal is not recovered, this system increases the time to COPS operation, provides cooling of the seals of the movable penetrations, and pmvides scrubbing of fission products in the drywell air space. Without ACIWA, the lower drywell would heat up after core damage and vessel failure until the passive flooder system actuates.
Flooder actuation will provide water to the debris in the lower drywell in a similar manner as the ACIWA system. However, the passive flooder does not add thermal mass to the containment, nor does it have the capability of mitigating suppression pool bypass.
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KHX= RHR heat exchanger effectiveness Ty , senice water temperature Containment sprays have a significant effect on the allowable steam bypass capability.
7 Use of sprays increases the maximum allowable bypass leakage by an order of magnitude and represents an effective backup means of condensing bypass steam.
b . ;j.f~6 ~, Q y 62.1.1.6 Suppression Pool Dynamic Loads During a LOCA and events such as SRV =ctmeinn, steam released from the primary system is channeled into the suppression pool where it is condensed. These actuation events impose hydrodynamic loading conditions on the containment system structures.
The containment and its internal structures are designed to withstand all loading conditions associated with these events. These hydrodynamic loads are combined with those from the postulated seismic events in the load combinations specified Subsections 3.8.2.3 and 3.8.3.3. A detailed description and definition of hydrodynamic loading conditions for structure design is provided in Appendix 3B. These loading conditions are briefly summarized in the following paragraphs.
6.2.1.1.6.1 LOCA Loads During a postulated loss-of-coolant accident (LOCA) inside the drywell, wetwell region will be subjected to the following three sequential hydrodynamic loading conditions of significance to structure design:
a Pool Swellloads a Condennrion NThrion (CO) loads a Chuggmg (CH) loads Following a pnemht-d LOCA and after the water is cleared from the vents, air / steam mixture from the divwell flows into the suppression pool creating a large bubble at vent exit as it exits into the pool. Bubble at vent exit expands to suppression pool hydrostatic pressure, as the air / steam mixture flow continues from the pressurtzed drywell. The ,
water ligament above the expanding bubble is accelerated upward which gives rise to pool swell phenomena lasting, typically for a couple of seconds. During this pool swell phase, wetwell region is subjected to (a) loads on suppression pool boundary and drag loads on structures initially submerged in the pool (b) loads on werwell gas space (c) impact and drag in2ds on structures above the inmal pool surface i
s.w cn . ..: syeaurs- Ameenment 31
6 6.2.1.1.5.g Suppression Pool Bypass During Severe Accidents The only mode of suppression pool bypass that presents any significant risk during a severe accident is vacuum breaker leakage. Vacuum breaker leakage results in the passage of gas from the drywell into the wetwell airspace. Vapor suppression and fission product scrubbing by the suppression pool are not available to the gas and vapor which pass through the vacuum breakers. The consequences associated with vacuum breaker leakage can be mitigated by the use of containment sprays.
Large amounts ofleakage can occur as a result of catastrophic failure of valve components or a valve sticking open. Lesser amounts ofleakage can results from normal wear and tear including degradation of the valve seating surfaces. For sufficiently large amounts ofleakage during a severe accident without containment heat removal, the time to COPS activation or containment overpressurization can be reduced and the amount of fission products released can be increased.
The probability that the vacuum breakers will leak or stick open will be minimized by using materials selected for wear resistance and using high quality seating surfaces. Additionally, the position switches which provide annunciation in the control room can sense a gap between the disk and the seating surface. If the gap is less than 9 mm, aerosols generated as a result of core damage can form a plug and terminate bypass flow. The severe accident analysis assumes the position switch can sense this gap.
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i drywell ventiladon system does not result in exceeding the design emironmental conditions for the safety-related equipment inside the containment, the dgwell system is not classified as safety-related.
6.2.1.1.9 Post. Accident Monitoring Refer to Subsections 6.2.1.7,7.2,7.3,7.5, and 7.6.1.2 for discussion ofinstrumentation r inside the containment which may be used for monitoring various containment
'Cf f[I, / C~ parameters under post. accident conditions.
(20' n 62.1.2 Containment Cubcompartments 6.2.1.2.1 Design Bases The design of the containment subcompartments is based upon the postulated DBA occurring in each subcompartment.
For each containment subcompartment in which high-energy lines are routed, mass and energy release data corresponding to a pocenbed double-ended line break are calculated. The mass and energy release data, subcompartment free volumes, vent path geometry and vent loss coefficients are used as input into an analysis to obtain the pressure / temperature transient response for each subcompartment. ,
62.1.22 Design Features The upper drywell, lower drywell and wetwell subcompartment volumes are covered in depth in Subsection 6.2.1.1. The remaining containment subcompartment volumes are:
(1) DryweB Head Region-The drywell head region is covered with a removable steel head which forms part of the containment boundary. The drywell bulkhead connects the RPV flange to the containment and represents the interface between the drywell head region and the drywell.
The DBA for the drywell head region is the double ended circumferential break of the 152 mm RPV head spray line of the CUW System at the connection to the RPV head nozzle. The other high-energy line in the drywell head region is the 51 mm main steam vent line. The RPV head spray line is chosen as the DBA for this subcompartment due to the higher mass and energy release rates from a postulated break of this line.
(2) Reactor Shield Annulus-The rmenr shield annulus exzsts between the reactor shield wall (RSW) and the RPV. The RSW is a concrete cylinder sunounding the RPV. The reactor shield wall is supported by the reactor pedestal and extends to a height 0.1m below the containment top slab.
6.2,26 Conteksment Syeenme- Amendment 31 l
4 Insert:
6.2.1.1.10 Severe Accident Considerations 6.2.1.1.10.1 Overall Containment Performance The containment structure provides for holdup and delay of fission product release should a core damage event occur. Core damage can only occur when all sources of core cooling are lost. Containment leakage during a severe accident is expected to be about the same magnitude as the allowable containment leakage.
Long term containment pressurization is governed by the generation of decay heat and non-condensable gases. The primary source'of non-condensable gas generation is metal-water reaction of the zirconium in the core. Non-condensable pressure buildup is accommodated by a relatively large containment volume and a high containment pressure capability.
The steam produced by decay heat is absorbed in the suppression pool resulting in a very slow containment pressurization and ample time for fission product' removal.
The limiting pressure bearing structure in the containment boundary is the drywell head. The Service Level C of the drywell head is 6.8 kg/cm2g This pressure capability is adequate to withstand the non-condensable gasses generated by reacting 100% of the zirconium in the core with water. The median ultimate strength of the containment is 9.4 kg/cm2g Ultimate strength capability is important for very rapid containment challenges such as direct containment heating and rapid steam generation. Evaluation of both of these phenomena indicate early containment failure is unlikely. Containment failure due to slower pressurization challenges are largely prevented by the Containment
- Overpressure Protections System as described in Subsection 6.2.5.2.6.
6.2.1.L10.2 Inerted Containment One of the important severe accident consequences is the generation of combustible gasses. Combustion of these gasses could increase the containment temperature and pressure. The containment will be inerted during operation to minimize the impact from the generation of these-gasses.
6.2.1.1.10.3 Lower Drywell Design The design details of the lower drywell are important in the containment l response to a severe accident. The key features are described below.
l
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23A6100 Rev.1 i t
ABWR stedsnr schnyAsdnit Report cnIXUon pressure at nominal temperature. The effect of temperature on the ur I dhk should be small, the analysis assessed the variability of about 2% per K (100 F .
f\
/ The area of the rupture disk is designed to permit the COPS system 20 be effective in l f mitigadng the pressure increase during an ATWS event in which th'$ operator controb 1 the injection flow. This provides ample raargin to steam genera [on rates related tu / [
decay heat generation. Analysis of the blowdown of the contaidIment following rupeh disk operation lodicates that the pool swell and the blowdo,wn loads will not threaten \ [
the piping, and that significant entrainment will not occur. ,
N / , :
This system is benefici(for several of the severe accidynt issues. In cases with continued '
core <oncrete attack, or 'those with no containment,6 eat removal operational, the t I containment will pressuriz'egThe COPS provides[ controlled release path preventing
- containment structural failur'e and mitigating f,twion product release. The COPS systerm dent behavior, e.g. debns coolability, in f reduces the the ABWR design. effect of uncertainties
\ in severe
\
/
/ f VesselDepressurization \,!
The ABWR reactor vessel is designed with a highly reliable depressurization system. TI- e l l nitrogen supply and battery capacity are sufficient to allow depressurization after RCIC failure during a long-term station' blackout This system plays a major role in preventirig f core damage. However, even irIthe event of akere accident, the RPV depressuruatif n i
f system can prevent the effefts of high pressure melt ejection. If the reactor vessel would fail at an elevated pressurb, fragmented core debris'could be transported into the uph j drywell. The resultingbeatup of the upper drywell could pressunze and fail the drywell.
Parametric an ' performed in Section 19AE of thehWR SSAR indicate that evbn
\ l in the event of, rect containment heating, the probability o l The RPV d,epressurtzation system 'lity further decreases the of this failure mechantsm.
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T,)fs details of the lower drywell design are important in the y_:..- w a;=~~ w ?- .:n L-v T- - - - --_ ; ' '4 response of he A]B j
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(N (1 Sacnficial Concrete i The floorof the ower drywell includ(a 1.5 meter layer of concrete D! above the containment liner. This is toansure that debris will not come in q direct contact with the containment boundary upon discharge from the b) reactor vessel.This added layer of concrete willprotect the containment from N possible earty failure.
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s importscr featurer identened byr she M PRA - Amerndmeert 31 0M
23A6100 M ,
ABWR sandort sawy Aurysis neport (2) Basaldc Concrete L
f The sacrificial concrete in the lower drywell of d.c ABdwill beowntruce l < $ of' low gas content concrete. The selection of concrete type is yet another l ' example of how the ABWR design has striven not only to provide severe ,
l accident mitigation, but to also address potential uncertainties in severe - .- <
~
t accident phenomena. Here, the uncertaintyis whether or not theforean be l ", '
cooled by flooding the lower drywell. For scenarios in which water fronuhi~ '
l
. ,.d lower drywell flooder is unable to cool the core debris, the concrete type ,
will result in a very low gas generation rate. This translates into a long g ggrw,,s ume to pressurize the containment. This is important because time is one of q w e. y 7 the key factors in aerosol removal.
9c w canx*w" e <A (3) Pedestal ,g g i t&
The AEWR'_ pedestal is formed.of"two concentric steel shells with webbing D I between them. The space between the shells is fil$d with concrete. The q' thickness of the concrete between the shells is L5S'm. A parametric study of x core concrete interaction was performed which indicated a very small
. potential for pedestal failure even in the event of continued interaction.
,) i Furthermore, any potential failure will not occur for approximately one day.
Q (4) Sump Protection 3 4d
'O Mm The lower drywell sumps are protected byten an shidi such that core debris will not enter them. This maximizes the upper surface area between the debris
. and the water and rnaximizes the potential to quench the core debris. The shields are made of alumina which is impervious to chemical attack from core-concrete interaction. The walls of the floor drain sump shield have channels O' which permit water flow, but which will not permit debris flow.The equipment I
drain sump shielo has no such channels. The height and depth of the shicids 4y has been specified to ensure that debris will not enter the sumps in the long q term. fuAkv dh m mon M% Ewe M 9u s %. A W i f. 2. . l . i. iDa4 has art arca E (5) Incr-~4oor Area i ga The floor area of the lower drywell has been maximi to improve tgh ? '/
potential for debns cooling. The lower drywell floor s$md88g . the gg a,omt
- Yr m f e,r' Lt WR Utility Requirements Document criterion of 0.02 . th.
yea., bd.%'5 (6) Wetwell-Drywell Connecting Vents M w d(s b c.c <c j a,y ,W W dob63 s p au.cL,g. 1
% sp< cad The flow area between the lower and upper drywell hW- d ri;;n-d hway />
or c eM* b% - ,
equatgventing of gases generated in the lower drywelt The 6 connectmg vents flow area is 11.25 m2 . This is important when considerin's a4
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, i 33A6100 C]ev.1 ABWR samtsrtsawysoms,ent neport 7ka nu ,cs da s p1 Aca-d amc4 cf eaer ~ &
loa:er dawcJi G pr:ce n s. yewad vesscJ Mhuc teau ti l Akt (wdckbr:s t%re y & inuta.:zd rak cf 1 nu&s s A lca)c/ cIrf)dI cverOc the
@etion rates associated with fuckoolant-interactions in the /**I h .
i lower drywell:The interconnection between the lower drywell and the wetwell N :
3 wit M [A ~ is at elevation -4.55,11.7 m above the floor of the suppression pool. Thus, cm J approximately 2.E6 kg of water must be added from outside the containment ""Cretcro 7
(
for the pool to overflow into the lower d W #' '
ell. g / g y*g [p/ *h 05be5 eht Win .J e w er o' ve '
g The path from the lower to the upper doweli inciudes severai vu cegree turns.vg
)I This tortuous path enables core debris to be strippedgrior to transport into the upper drywell minimizing the consequences from high pressure melt 4
4 ejection. Aho important when considering high pressure core melt scenarios, ah Q the configuration of the connecting vents will result in the transport of some
,s' core debris directly into the suppression pool. This is preferable to transoort into the upper drywell and would result in the debris being quenched with ;
only a slight increhe in the suppression pool temperature.
U (7) Solid Vessel Skirt,7~)"
eow Jmson
-e cA W das O- y Jeo.G s-
\3 ./ A 3c.c a cci d ovco S The vessel skirt ' " ^ n does not have any penetrations which would allow thi flow of water from the upper drywell directly to the lower drywell. ;
This f nsures a very low probability that water is in the lower drywell before the !
/ time ofvessel failure. Thus, large scale fuckoolant interactions are precluded.
(oec ^ " S.n.;; -
bn' gasses of mbustion the important severe accident consequences of these gasses could increase the containment temperg d andis the genel pressure. GWR containment will be inerted during operation to 'mize the l impact from the neration of these gasses.
ContaintnentIsolatio ,
The ABWR containment demgn has striven to minimir e number of penetrations. '
- This impacts the severe acciden . ponse due to maller probability of containment .
solation failure. Alllines which origt tein reactor vessel or the containment have i l 3
ual barrier protection which is generd .ained by redundant isolation valves. Linest ,
j hich are considered non-essa ' m mitigau an accident isolate automatically in f !
r sponse to diverse isolario I gnals. Lines which gbe usefulin mitigating an f a cident have means t etect leakage or breaks and may isolated should this occun U ded Pressure Pl> lng low essure piping in the ABWR has been upgraded to withstan er pressure. ,
Weduces the probability of an interfacing system LOCA and the severe ident 7,consennenea
" *b --h un event. 1 Importerrt Foerures Identdsed by the ABWR PRA - Amendment 2y T & 25
M i
saa m ca.s ;
ABWR swent seerpaania normer i
6 2.t. . m.y 39ED' Corium Shield 49 EDA-iseud During a hypothetical severe accident in the ABWR, molten core debris may be present on the lower drywell (L.D) floor. The EPRI ALWR Requirements Document specifies a floor area of at least 0.02 m2 /MW t3 to promote debris coolability. This has been interpreted in the ABWR design as a requirement for an unrestricted LD floor area of 79 m 2 The ABWR has two dram sumps in the periphery of the LD Door which couki collect core debris during a severe accident ifingression is not prevented. Ifingression occurs, a debns bed will form in the sump which has the potential to be deeper than the bed on the LD floor. Debris coolability becomes more uncertatn as the depth of a debris bed mcreases.
The two drum sumps have different design objecoves. One, the floor dram sump, is designed to collect any water which falls on the LD floor. The other, the equipment drain sump, collects water leaking from valves and piping. Both sumps have pumps and instrumentation which allow the plant operators to determine water leakage rates from various sources. Plant shutdown is required when leakage rate limits are exceeded for a certain amount of time. A more complete discussion on the water collection system can be found in Subsection 5.2.5.
19ED.2 PemM Dr!M wit!
A protective layer of refractory bricks-a corium shield add be built around the sumps to prevent corium ingression. The shield for the equipment dram sump *ould" b
-b/ solid except for the inlet and pudet piping which would gM-h it: roof. The shield for the floor drain sump -dJd bc similar except that it mbu channels at floor level to allow water which falls onto the LD floor to Dow into the sump.The height of the channels $31 bc chosen so that any molten debris which reaches the inlet will fr e before it exited and spilled into the sump. The width and number of the channels be chosen so that the requtred water flow rate during normal reactor operation is achievable. A i: :h ef A eq for idivu. bm ,omp ii:Id i: ira. in Hgure C ED-1. 'l Th of the sump shiel[(,vud Eddfonly have to be thick enough to withstand ablation, if any is the chosen wallmaterial.
The walls of the floor drain sump shield u..=1._ .._.d) must be thicker so that molten debris Dowing through the channels has enough residence time to ensure debris solidiScation. l Both shields extend above the LD floor to an elevanon greater than the expected mmmum height of core debris. Thus, no egmficant amount of debras will collect on conum snow-Amendment 21 1)AN
. ~ . -- .. .. . . - . - .. .-
7I ;
. nassoow. r l ABWR smssenyAns&seRepet l the shield roofs. The solid shield can be placed directly on top of the LD floor. The !'
channeled shield Eld.;ca refractory bricks embedded into the LD floor beneath the shield to prevent core-concrete interaction involving the molten debris in the channels.
C
--- -. W i Tlie analyses presented in Subsections 19ED.4 and 19ED.5 provide a basis for sizing the > ;
proposed design of the floor drain sump corium shield. j
.39EDI Success Criteria s Timsd On:;;;n E l
For the proposed design to be considered ma=Jul, it must satisfy the following [
requirements: ;
i (1) Melting Point of Shield Matenal Above Initial ContactTemperature {
1 The shield wall material shall be chosen so that its melting temperature is [
greater than the interface temperature between the dehns and the shield wall.
(2) Channel Length j The length of the channels in the shield must be long enough to ensure that a plug forms in the channel before debris spills into the sump. The freezing 1 process is expected to take on the order of seconds or less to complete. l (S) Shield Height, H ,u Above Lower Dr7well Floor The shield height above the lower drywell floor shallbe chosen to ensure long i term debris solidi 6 cation. The freenng process will be complete during the l time frame when the shield walls are behaving as senu-infinite solids. In j f
addition, the shield must be tall enough to prevent debns from accumulating on the roof of the shield.
t (4) Shield Depth. Hw Below lower Drywell Moor !
l The shield depth below the lower drywell floor shall be chosen to ensure long j term debris solidi 6 cation. !
}f4 WaterHowRate M The total flow area of the shield channels shall be great enough to allow water i
j flow rates stated in the Technie=1 Specifications without causmg excessive -
water pool formation in the lower drywell.
{g) Chemical Resistance of Shield Walls ;
& The wall matenal chosen for the corium shields must have good chemical I
t resistance to susceous slags and reducmg envuonments. W=ne- can be ;
i
/
rg - cown musw-Amenement27 f
t
,1 I
23A8100 Cev.1 ABWR saadantsarayanarysisnoreer determined to a first degree by comparing the Gibb's free energy of the oxides ,
which make up the shield wall and the oxides present in core debris.
(7) Seismic Adequacy The seismic adequacy of the corium shields will be determined in the detailed design phase. Adequacy should be easily met because the shields are at the lowest point in the containment. Missile generation is notan issue because the shields are not near any vital equipment.
.~
Subsection 19ED.6 contains an example of success calcnh6an= for requirements (1)/
through (4) for a chosen channel height ofI cm.
19ED'.4 Analysis of Shield Freezing Ability /
Heat transfer and phase change analyses are presented in this subsectiotr to determine the'fcasibility of a channeled shield to prevent molten debris ingression into the floor drain samp. Two time frames were considered. First, a freeze front ' analysis was performedfor early times (seconds or less) to determine the time required to form a plug. The Ioniperm ability of a plug to remain solid was determined using a steady-state analysis.
\ ,
19ED.4.1 Assumptions j
\
The major assumptions invoked in the analyses and their bases follow:
s (1) Molten debris enters eeh=nnel with negligible superheat.
Molten debris interacts wi material (steel, concrete, etc.) and the N
lower drywell environment as it passes from the vessel, contacu the LD fbor and spreads to the shield. This intAraction depletes the molten debris of any superheat and can result in eutectic fhr; nations.The melting temperance of core debris which has undergone little imeraction is approximately 2500 K.
Significant interaction with the concrete flo'br reduces the debris melimg temperature to approximately 1700 K. 'NNg
/
(2) Duringde freezing process, the temperamre profile \of the solidifie rapidly obtains its steady state value.
This assumption introduces little inaccuracy because:
(a) the heat conduction coefficient in the solidified deteis is Pntly
/ larger than that of the shield material.
(b) the depth of the solidified debris is consaderablyless than the height of
/ the shield. \
x w m.w a - - n as~
Y '
./ ,
\
N 6.2.1.1.10.4.2 Corium Shield Design The corium shields are constructed of alumina. The height of the shields above the floor is 0.4 m. The floor drain sump has channels I cm high. The channel length must be at least 1.06 m. The depth cf the channeled wall below the floor is 0.4 m .
6.2.1.1.10.4.3 Design Evaluation Alumina has a melt point which is grester than the contact temperature of the core debris. It is resistant to reduction reactions with the metals which make up core debris. The height of the shields meets the requirements for to ensure long term debris solidi 6 cation and to prevent material from accumn12 ting on the roof of the shields. Similarly. the depth of the channeled wall will ensure long term debris solidification. The height and length of the channel for the floor drain sump will ensure debris freezing. ,
The details of the analyses leading to these conclusions may be formd in Attachment 19ED to Appendix 19E.
Markup co M c leele &
23A6100 R:V1 rew re a eca'de ,t e
(
is rr'y 4 r y ptDmlcl pYReport eft 9ffJ
' ABWR Stenbrd Safety Analys m;mjby 3:// Ta k~ ,
l (3) The ECCS flows assumed available are 2 HPCF, I RCIC, and 2 LPFL (RHR). !
(4) Containment cooling is iniated after 10 minutes (see Response to Quesdon 430.26).
Analysis of the net positive suction head (NPSH) available to the RHR and HPCF pumps in accordance with the recommendations of Regulatog Guide 1.1 is provided in Tables 6.2-2b and 6.2-2c, respecdvely.
General compliance for Regulatory Guide 1.26 may be found in Subsection 3.2.2.
6.2.2.3.2 Summary of Containment Coofing Analysis When calculating the long-term post-LOCA pool temperature transient,it is assumed that the initial suppression pool temperature and the RHR service water temperature are at their maximum values. This assumption maxmuzes the heat sink temperature to which the containment heat is rejected and thus maximizes the containment temperature. In addition, the RHR heat exchanger is assumed to be in a fullyfouled condition at the time the accident occurs. This conservatively minimizes the heat exchanger heat removal capacity. Even with the degraded conditions outlined above, the maximum temperature is maintained below the design limit specified in Subsection 6.2.2.1.
It should be noted that, when evaluating this long-tenn suppression pool transient, all heat sources in the containment are considered with no credit taken for any heat losses other than through the RHR heat exchanger. These heat sources are discussed in Subsection 6.2.1.3.
It can be concluded that the consenative evaluation procedure described above clearly demonstrates that the RHR System in the SPC mode limits the post-LOCA containment
.1719 e/C 6.),7 7,3 ---- temperature y transient.
6.2.2.4 Test and Inspections The Containment Cooling System (CCS) is required to have scheduled maintenance.
The system testing and inspection will be performed periodically during the plant normal operation and after each plant shutdown. Functional testing will be performed on all active components and controls. The system reference characteristics will be established during preoperational testing to be used as base points for checking measurements obtained from the system tests during the plant operation.
The preoperational test program of the CCS is described in Subsection 14.2.12.The following functional tests will be performed. The RHR pump will be tested through the suppression pool cooling loop operation by measunng flow and pressure. Each pump will be tested individually.
ComsinmentSystems- A.T nd.i at31 6.2-33 j i
i
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400d /W' t 8.2.2.3.3 Severe Accident Considerations ,
The containment spray features of the RHR System can reduce the amount of radioactive material released to the environment in the event that core damage occurs. The benefits provided by the sprays are condensing steam, scrubbiag of fission products in the containment !
airspace, and supplying water to ex-vessel coro debris. The conditions for activation of the containment sprays are described in the En ergency Procedure Guidelines in Appendix 18A. .
The water sprayed into the upper drywell absorbs heat from the RPV outer surfaces and the debris which relocates into the upper drywell, if any, -
upon vessel failure at high pressum; Cooling of the upper drywell prevents any potential for overtemperetare failure of seals in the large operable penetrations (e.g., the drywell head, equipment hatches and personnel airlocks). Water which collects on the upper drywell floor is direc%d into ,
the wetwell thmugh the connecting vents.
The containment sprays provide significant mitigation of suppression pool bypasa. The incoming water absorbs heat and condense steam. While the heat absorption is not as efficient nor as extensive as what would occur if ,
the suppression pool was not bypassed, the time to COPS activation or containment failure can be delayed significantly. This delay results in a i sigmficant reduction in the radioactive release due to fission product -
decay, t
The water sprayed into the containment also scrubs fission products !
which are in the containment airspace. Scrubbing reduces the amount of .
radioactive material which is available for release from the containment. [
.?
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See+ert Accdeer c rit.rj/1 c7
' 23A61MW 1 A f &cr7a (el ABWR 70$ntsareryAnairsis stra soport9f,7c 6.33.7.8 Bounding Peak Cladding Temperature Calculations Consistent with the SAFER application methodology in Reference 6.3.' the Appendix K peak cladding temperatures calculated in the previous sections must be compared to a statistically calculated 95% probability value. Table 6.3 6 presents the significant plant variables which were considered in the determination of the 95% probability PCT or the i sensitivity study. Again, since the ABWR LOCA results have a large margin to the '
acceptance criteria, a consenutive PCT calculadon was performed which bounds the 95% probability PCT. This bounding PCT was calculated by varying all plant variables in the conservauve direction simultaneously. The results of this calculanon for the limiting case are given in Figures 6.3-67 through 6.3-75 and Table 6.3-4. Since the ABWR results have large margins to the 10CFR50.46 licensing acceptance criteria, the ABWR licensing PCT can be based on the bounding PCT which is well below the 1204*C PCT limit.
633.8 LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2,it is concluded that the ECCS will perform its function in an acceptable manner and meet all of the criteria in Appendix 4B, given operation at or below the MAPLHGRs provided
% ferT, by the utility for each fuel bundle (Subsection 6.3.6).
)
f J.3 T6.3.4 Tests and Inspections
/). S 63.4.1 ECCS Performance Tests
'I:#r All systems of the ECCS are tested for their operational ECCS function during the fAfg preoperational and/or startup test pmgram. Each component is tested for power
' source, range, direction of rotation, setpoint, limit switch setting, torque switch setting, q3 etc. Each pump is tested for flow capacity for companson with vendor data. (This test is S./
also used to verify flow measurmg capability).The flow tests involve the same suction and discharge source (i.e., suppression pool).
Alllogic elements are tested individually and then as a system to verify complete sptem response to emergency signals including the ability of valves to revert to the ECCS alignment from other positions.
Finallhthe entire system is tested for response time and flow capacity taking suction f rom its normal source and delivering flow into the reactor vessel.Thislast series of tests is performed with power supplied from both offsite power and onsite emergency power.
See Chapter 14 for a thorough discussion of preoperational testing for these systems.
See Subsection 6.3.6.2 for COL License information regarding ECCS testing requirements.
I 6.3-20 EmWQency Core Cookng Symanme - Amenamt 31
.- l i
6.3.3.9 Severe Accident Considerations In the unlikely event that the ECCS does not prevent core damage, its operation (recovery if necessary) can be beneficial in mitigating the j consequences of core damage. The analysis of core damage events was )
performed using best-estimate methods rather than design basis codes !
such as SAFER /GESTR.
The primary injection path for the RHR system during a severe accident is into the vessel via the LPFL header. The conditions under which the LPFL should be used are described in the Emergency Procedure Guidelines, see Appendix 18A. For injection to occur, the RPV must be at low pressure.
If the LPFL is not initiated in time to prevent core damage, its use can mitigate the consequences of core damage by enhancing cooling and preventing radiative heating from the core debris. Ifinjection is initiated prior to vessel failure, melt progression may be arrested in-vessel.
However, if vessel failure occurs, debris will relocate from the vessel. After vessel failure, water injected into the vessel will flow out of the vessel breach into the lower drywell. Water flowing into the lower drywell will cover the core debris and enhance debris cooling.
l l
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23A6100 Rev.1 g g ,
ABWR seemtantsarery Amstysis neport b >
j?e znede df%al B!yd()
Site acceptance testing, periodic surveillance testing and preventive maintenance, l inspections, etc., shall be performed in accordance with the manufacturer's recommendations, including time intervals for parts replacement, the plant ;
maintenancne program, and the operational reliability program (see !
Subsection 9.5.13.19) -
'i 9 5.11.5 Instrumentatiort Requirements The CTG is provided with local instrumentation and control svstems suitable for ,
manual startup and sh tdown, and for monitoring and control during operation. '
sof '
Automauc startup an(d*'_Y*g** *#f 1i s controlled via the control console located m-t
- the main control room. Controls are also provided in the main control room for manual I
startup of the CTG, and to facilitate connections to the Class IE buses, should a station hlackout occur. ,
m , ,
. i _ _ _: a :
' ' ' hatrol room displays are provided -
to monitor starting, lubricating and fuel supply systems, the combustion air intake and exhaust systerp, and the excitation, voltage regulation and synchronization systems.
CT& sikT(step cafabrInig n fronded in tore mion ecM ee f t o e 'e-o . ;
Generator output voltage, current, kVA, power factor, Hz, etc., are also displayed in the control room. Annunciators and computer logs provide early detection of abnormal .j behavior.
9.5.12 Lower Drywell Flooder ;
9.5.12.1 Design Basis ,
The function of the lower dnwell flooder (LDF) is to flood the lower dnwell with water ,
from the suppression pool in the unlikely event of a severe accident where the core melts and causes a subsequent vessel failure to occur.
The equipment shall meet the following performance criteria:
3 (1) The LDF shall provide a flow path from the suppression pool to the lower i
drywell when the drywell air space temperature reaches 2f>0*C.
ud proids 0.A equg yot>l of 4ctW .~ %
(2) The LDF shall pass suffic' nt flow from the suppression pool to the lower [m drywell to quench all of e postulated corium, cover the corium, andkemme the corium decay heat, as confirmed by severe accident analysis i (App:+ !9Ek (Sasenos IW 2.12.). j 1
(3) The LDF shall operate automatically in a passive manner. .)
(4) The LDF outlet shall be at least one meter above the lower dnwell floor.
Otheran=.E==y Syaeanne-karnderment 31 9.5-61 l
1 l
- - - - -- - -= - -- -
s Qr
- 23A61 llev.1 ,
ABWR stuhntseroty Anar sis r neport (5) The LDF inlet shall be located as far below the bottom of the first horizontal dnwell-to-wetwell vent as possible while still meeting the requirements for the location of the LDF outlet.
(6) The LDF shall not become a flow path from the suppression pool to the lower dnwell during design basis accidents (DBk.) such as loss of-coolant accidents (LOCAs) or during normal plant operation.
(7) The LDF shall distribute flow eveniv around the circumference of the lower dnwell.
B f.j j 9.5.122 System Description
,3 g .
] 5g}4
~
bw 3 3 The LDF, shown schematically in Figure 9.5-3. provides a flow path for suppression pool water into the lower dnwell area during a severe accident scenario that leads to core h]jfj# N' meltdown, vessel failure, and deposition of molten corium on the lower drywell floor.
Molten corium is a molten mixture of fuel, reactor internals, the vessel bottom head and
%,I dd control rod drive components. The flow path is opened when the lower drywel m orms a. pod of N s(
5i J~ cN temperature reaches 260 C. o, den, we. cerc dtbru .
9 3- 1 >hy :- - s poe cootss *i d q The flow of suppression pool water to the lower drywell through the LD p. . )ne molten corium and subsequently removes the corium decay heat.This limits the dnwell
$j?$cj' $1'N temperature to 260'C and avoids degradation of non-metallic penetration seals in the S}f}Q g g .y upper and lower dgwell. Interaction between corium and the concrete floor is also stopped. This delays the time of fission product releases for the severe accident, which i A
allows for more decay of fission products and results in lower releasde f ctions.M b"" ?"M 9 Fe 1,ag ji ecacaud kom W dstm bed 4 be. scrubbed og.W o A y w cler po d . I Y,$)M,::- The LDF consists of ten pipes that run from the vertical pedestal vents into the lower !
3 -M t 3-[ drywell. Each pipe contains a fusible plug valve connected to the end of the pipe that l 2
g 'd4 extends into the lower dowell by a flange. The fusible plug valves open when the dowell l Qjqs air space (and subsequendy the fusible plug valve) temperature reaches 260*C. When ;
i j_dj the fusible plug valves open, a minimum of 10.5 L/sec of suppression pool water will be ,
.5
- W d 9-$ '5 supplied through each flooder 'pe (105 L/sec total) to the lower dnwell to quench g' d the corium,ha c=rt %... wewam
-. . .T. remove corium decay heat, which is estimated at 1%
, e ;
M of rated thermal power. The flow rate is based on a minimum hydrostatic head of 200 j f )k~}gd "3 J mm above the flooder pipe inlet centerline and takes the frictionallosses through the l
% flooder pipe and fusible plug valve into account. [
The inlet centerlines of the drywell flooder pipes are located 10.2 meters below the ;
bottom of the vessel, rnd the outlets of the fusible plug valves are located at least one. l meter above the lower dnwell floor.
% d*WN suppcc u ocnO Peoi xaM Sre 1 Cor m wtA 4 le m ucr.a. l The fusible plug valves are made from flanges welded to e end of the ttnt ms e the lower drywell area. The inner diameter of the pipe is slightly en ed to accommodate ,
a stainless steel separaticn disk, an insulating disk and fusible metal. e insulating disk V5-62 Other Auxiliary Syseems - Amendment 31
i' 23A6100 Rev.1 ABWR studard safety Analysis Report Tci h n .;aic a Er k ,,n.J.em dx Cd ca..Lu rt nM u .5 ttfL
- h o Q &cw.y kb. : t- 'T Y v.d a ,s.ma 3,;kuch 9<& n %p.v.m ~
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- Q ce rei cr A., us c abcx r*4 p! o m . .y thermally insulates the fusible metal from7 wetwell water to assure that the fu ble e r v.a u metal is not cooled by wetwell water and prevented from melting during the severe Cr"" %
accident high lower dnwell temperature conditionT. The end of the fusible plug7 vahe"?'i " -
n . c, n xm is cmered with a plastic cover that has a low melting point. The purpose of the cover is J ,.a ,
to avoid corrosion of the fusible metal material and to assure that any toxic components % 3,9 from the fusible metal material that might be released do not escape into the lower a .et N S ^ Md dt3well area during normal plant operation. m t : ., ac .
The fusible plug valve is mounted in the verucal position, with the fusible metal faciIg '* ? 'd downward, to facilitate the opening of the valve when the fusible metal meltin{'
temperature is reached. k m4up h b & dca bd*
a su.s & mm.au &cm W sk M 1.ca dig:.A 3a spca.r. reaura LL The dnwell flooder pipes are welded to the stainless steel vertical vent pipes in the pedestal and to the steelliner in the lower drywell.
9.5.12.3 Safety Evaluation 9.S.i?. 3.l G cm2A F_.aaaben
- The LDF is a passive injection system and is maintained in an operable state whenever the reactor is critical. The system is never expected to be needed for safety reasons because of the extensive array of water injection systems available to maintain core cooling.
l The LDF is safety-related because it is a structural extension of the blowdown vent system. The LDF is Seismic Category I. The quality control classification of the LDF components is the same as the pedestal and the blowdown vents. Therefore,it meets the same structural design, matenals, welding, fabrication, thermal and structural analysis, 1 and quality assurance requirements as the reactor pedestal.
The LDF has sufficient redundancy that the failure of one fusible plug to open does not degrade the ability of the system to flood the drywell and quench the corium.
2 The design pressure of the LDF component is 1.1 kg/cm d.
The design temperature of the LDF components is 171*C.This value is the priman-containment design temperature and considers DBA events. If the LDF components lose pressure integrity at higher temperatures during a severe accident, then the LDF !
function (i.e., dryw:ll flooding) is performed. Therefore, the design temperature does not need to be higher than the temperature based on DBA events.
The LDF components have zero leakage when subjected to design differential pressure of 1.1 kg/cm 2at a design temperature of 171*C.
The portions of the flooder pipe that extend from the steel liner in the lower dnwell meet the requirements of ASME Class 2 piping components.
Other Auswery Spnems- Amendment 31 9.543
- 23A6100 Rsv.1 ABWR- standardsafery Anotrsis neport An ANSI B16.5 stainless steel weld-neck flange (or equivalent) is used at the interface i betw6en the flooder pipe and the fusible plug valve. The flooder pipe is made of the
! same material as the blowdown vent pipe or of a stainless steel material that is compatible for welding to the blowdown vent pipe.
The fusible plug is required to open fullywhen the outer metal temperature of the valve reaches 260 C during a severe accident and to pass a minimum of 10.5 L/sec with 375 mm of water above the uhe inlet.
A plastic cover on the valve oudet se ils the valve from the intrusion of moisture that could cause corrosion of the fusibir ,etal material. The plastic cover has a melting point below 130*C and greater than 70 C and is required to melt completely or offer minimal resistance to valve opening when the opening temperature is reached.
IM ms h+ (610 & 4 9.s. ta 3,a W %.5,60.3 3 9.5.12.4 Testing and inspection Requirements The ability of the LDF to mitigate severe accidents by passing sufficient water to cover and quench the postulated corium in the drywell is confirmed O'"' bv.PitX
- anag (Appendix L90).
14C No testing of the LDF sptem will be required during normal operation. During refueling outages, the following surveillance would be required:
(1) During each refueling outage, verify that there is no leakage from the fusible plug valve flange or outlet when the suppression pool is at its maximum level.
-hoo (2) Once everyfourrefueling outages, lower suppression pool water level or plug the flooder pipe inlet and replace two fusible plug valves. Test the valves that-were removed to confirm their function. This practice follows the precedent set for inservice testing of Standby-Liquid Control System (SLCS) explosive valves in earlier boiling water reactors.
9.5.12.5 Instrumentation Requirements The LDF operates automa6cally in a passve manner during a severe accident scenario that involves a core melt and vessel failure. No operator action is required; therefore, no instrumentation is placed upon the system. An inadvertent opening or leak muld be detected by the lower drywellleak dr. tion system and the suppressica poc oer j I level instrumentation which would rese in plant shutdown.
l l
During severe accidents, operation of the LDF is confirmed by other instrument readings in the containment. These instrumw.s include those which would record the ,
I drywell temperature reduction and the lov'enng of suppression pool water level.
I l
Other Auriliary Systems - Amendment 31 9.5 64
_ - - __-__ - __ a
- mem w. ,
ABWR saadantsarnyA=sysis aman
\ /
Flooder sctuation is expected to occur approximately five hours after reactor scram during mos'tsevere accident scenarios. The decay heat level ac 'this time is approxim ately 1% of the rate' d power. Assuming the entire core relocates to the lower drywell, the 2 debris bed will h' ave a decay heat generadon rate, Qg,of 39 MW. If all of this her.t is ;
transferred to the fl9oder water, the rate and time to fill the lower drywell arc l
./3 frill.d = OA86 /sec l
\
N'
- 21 minutes t ,
N' / /
The maximum heat flux from th7 surface of a debris bed that has been experimentally
- observed (Subsection 19EB.2.37is 2 SW/m 2. The lower drywell has a surface area of 88.25 mr .Thus, the maxunurn cooling rah of the debris bed. Q ,is 177 MW. For this heat transfer rate, the rate'and time to S11Ihe lower drywell are N
8 frill, max = 0.016'tq /sec
/ t all. max = 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> . ,
ed as the debris is In pracpe, this high heat flux is not expected to be main '
quenched. Nonetheless, the time to fill the lower drywell to the wievation
' of the flooder exit'will be bounded by these two values,21 minutes and 1.8 hottrs. This difference in +
dning will not have a significant impact on the fission product rele from the contamment since the steam produced during debris quenching will f anyEssion products released during this time into the suppression pool.
QK 9 5. l Q.3,2of One Flooder Une Opcning First Consequences Core debris that enters the lower drywell will be distributed fairiy uniformly. The lower drywell floor was designed so that debris spreading would not be hindered. The temperature of the lower drywell air space and structures should be even more uniform because of convective and radiative heat transfer from debris material. Cooler regions will tend to absorb more heat than warmer ones resulting in temperature equalization. .
J However, if highly non-uniform debris dispersal occurs, it has been posenkred that one flooder line could open and its operation could delay or even prevent the other lines '
from activating. In the worst physical case, the initiation of one flooder line causes crust formation without completely quenchmg the debris. The crustlimits heat transfer from the surface of the debris bed. Core-concrete interaction (CCI) will occur if surface heat ,
transfer is reduced enough.
4 23A610tl Rev.1
. ABWR sentnisaferyA=tysis noron CCI results in large quantities of gases being formed under the surface of the crust. The gases willincrease in pressure due to continued generation until the crust ruptures or they escape from the edges of the bed. In either case, the gases will pass from the debris l bed into the lower drywell airspace. The passage either will be unobstructed with gasses exiting the deb-is above the water elevation or through an overlying layer ofwater. Since .
only one flooder line is presumed active, the water layer, ifit exists, will be thin and no significant amount of heat will be transferred from the gas to the liquid. j Concrete has an ablation temperature of approximately 1500 K. The released gases from core concrete interaction will be at least at this temperature. IIigher tempe:atures may be reached by the gases as they interact with debris matenal in their exit. Thus, gases enter the lower drywell air space at very high temperature. The CCI gases will increase the temperature of the lower drywell air space. More flooderlines will become l active as the lower drywell temperature increases. For this reason, the activation of a single flooder line is transient condition atworst and is not expected to adversely affect ,
the operation of the other lines.
\
19E.2.8.2. ve Opening Time / l The filsible plug valve is designed to open when the lower drywell temperature reaches 533 K. Tlie usible matenal is made up of an alloy mixture of twqor more of the following me : tin, silver, bismuth, antimony, tellurium, zing'and copper. Alloy contents are chosen so that the plug melts when its temperafure reaches 533 K.
/ .
,c' [
~
The melting points of individual metals are as follows:
Met Meftkng Point (K)
Antimony (Sb) \x / 903 Bismuth (Bi) 7
/ 544
- N Copper (Cu) 'L356
\
/'/ 1233 k
Siher/Ag)
Tdilurium (Te) 722
./
,- Tin (Sn) 505
/ Zine (Zn) 692 \
The basic configuration of the fusible plug valve is shown in Figure 19E.2-24. lastic l cap has a melting point much lower than that of the fusible plug. Flow initiation occurs
' i camaewww. at we pararma.-n .-m 21 yz:W l
i 22As100 nov. t
. ABWR suunntseteryAnasysisasper 5
when the small aimular groove,2.0 mm in depth, melts. Hydrosta c pressure then expels the. remainder of,the plug, the stainless steel disk an the teflon disk.
The valve opening time is the, time required to melt p the fusible metal in the annular groove. To estimate the opening time, a calculation p has been made for a pure bismuth plug. Bismuth was used because it has.the closest melting point to 533 K.
A' Heat transfer from the surrounding 1Itainless steel pipe to the plug is by conduction.
Heat transfer from steam in thMower drywell to the stainless steel pipe is by convection.
The pipe also receives radiative heat from the debris on the lower drywell floor. Heat transfer to the bottorrtof the valve was neglected. Th'edebris bed surface temperamre and lower drywell gd temperature were estimated using'a representanve MAAP-ABhR sequence. Using'these assumptions, the valve opening time'was calculated to be less than appp*im= rely 10 minutes depending on the steam absorbtivity. This is a repregntative time from when the lower drywell gas space reachSNi33 K until th,:
flp/ der line becomes active. \ N
% Cl.S.62.3 3
-198;2.82J Estiniation of Net Risk A stodos i%212.4 \.au at' A In order to ese the net risk of the flooder sys a sensithity study was f lowes J performed using three failure probabilities for the enM-n e ~n t m2 In these cases, the failure probability odergeg of the M, in, d ooderwas increased from its base case value of 0.001 to 0.01,0.1, and 1.0. . .
pu%o,o 64 W Ase,dey'5j not sensitive to ei+r ;r.ctcr. 4 N uu .
As ind'cated in Tol@? r 10E.2= 2 the overall Failure of the pr.== ,.=de:-leads to an increase in the probability of Dry CCI. Thus, WO the probability of Dry CCI increases by one, two and three orders of magnitude, respectively for the three sensitivity cases. However, the base case results for Dry CCI are so small that a three order of magnitude increase does not impact other results ,
significantly.
The principal conclusions of the sensitivity studies are:
(1J Pedestal failure does not increase since it is dominated by 6: Mt GGIA sequencess whare. con enuck, i.duacWoo pus.sw ch A$
ha Wee 69d b Moodul.
(2) The only probabilistic output which shows an significant vanation is dryvell head seal overtemperature leakage Gb O. which ewhihits a twofold increase for a two orders of magnitude increase in the passive flooder failure i probability, and a ten-fold increase for a three order of magnitude increase.
The change in seal leakage is much less than the change in passive flooder 1
failure probability since high RPV pressure sequences with entrainment of debris to the upper drywell and failure of the upper drywell sprays dommate the seal leakage sequences in the base analysiz.
1stz$12 onaminietk Anewe amant Perkmance - Amendment M I
23A6100 Rev.1
.' ABWR sumtentserayAantysmaoper (3) ven for the case where the passive flooder is assumed to be unavailable, the 4 associated with the Dry CCI iWy '.5E-10. Since only the Dry CCl cases have failure of the passive flooder, this frequency represents an upper j nund for the imp _agt_of passive flooder failure on offsite dose, s benea h %v h+t censidued A h mjuchen ef etL te dcec .
Tpus, it is s en that the lower drywell flooder does not affect net dsk for frequencies M ^ Therefore, ac ch:rt of the ir. pat en dsk c.s w Md7h value of the lower drywell flooder system is not measured as a direct impact on risk. Rather, it should be viewed as a passive system which serves to limit the impact of uncertainty in operator actions and allows the ABWR design to mitigate a severe accident in a purely passhe manner.
19F 7 R}.4 Summary apassive flooder meets its design goal of preventing or, at least, mitigating core concrete interaction in the lower drywell.The flow rate required,so remove the heat generate'd.in the debris bed is 0.018 m8/sec which can be provgied by two of the ten flooder linh The expected flow rate is 0.099 m8/sec (nine ofthe ten lines active). If the expected flow is achieved, a one-meter layer of waterfeill be established above the bed in a time be en 21 minutes and 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after fidw initiation. One flooder line opening first is not ex ected to prevent the other linch from opening dudng a severe accidentin which si cant amounts of core debrisiis present in the dqwell. The flooder lines will become tive within ten minuths of the lower drywell gas space reaching 533 K. The passive Hooder has negligible impact on the net risk of the plant since it provides a redundant fhetion to the firewater addition system.
19E.2.8.3 Corium Shield During a hypothetical severe accidedt in ies ABWR, molten core debds may be present on the lower drywell floor. The EPRI AL Requirements Document specifies a floor I area of atleast 0.02 mr /MW promote de coolability. This has been interpreted ;
in the ABWR design as a regmrement for an unrestricted lower drywell floor area of 79 m r,- 7 The ABWR has two sumpsin the penpheryof the er drywellfloorwhich could collect core debris ' g a severe *Mentifingrenian is et prevented. Ifingression ;
occurs,a debris willformin the sumpwhich has the po to be deeper than the bed on the lower drywell floor. Debris coolability becomes more certam as the depth l of a debds bed increases.Therefore, debris should be kept out ofthe sumps. !
The two tr/ have different design objectives. One, the floor ain sumps (HCW) sump,,pollects water which falls on the lower drywell floor. The other, th ipment draig'(LCW) sump, collects water leaking from valves and piping. Both s have l pp[nps and instrumentation which allow the plant operators to determine e,b ge
- 1 1.9E.ZJs3' Determmistic AnsMis of Ptant Per%rmance - Amendment 31 /
~
'j t a > .. a 23A6100 Rsv.1 ABWR . standantsafety Analysis Reper: i
'E
- " *E t l v[RT[ CAL STEEL LINER .* p t. U N .
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Au diawnsions la nAHauters. .!
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i Figure 9.5-3 Lower Drywell Flooder System Arrengement/ Configuration .
6
~ 9.5 79 Other Auxiliery Systems - Amendment 31
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