ML20115G220
| ML20115G220 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/10/1996 |
| From: | Quirk J GENERAL ELECTRIC CO. |
| To: | Grimes B NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| MFN-079-96, MFN-79-96, NUDOCS 9607190012 | |
| Download: ML20115G220 (149) | |
Text
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GENuclearEnergy Joseph F. Quirk GeneralBectnc Company ABWR Licensing Manager 175 Curtner Avenue. wC 782; San Jose, CA 951251014 ProjectManager ABWRCerticaten 408 925-6219(phone) 406 9254257(facsimile)
June 10,1996 hf FN 079-96 Docket No.52-001 Document Control Desk U.S. Nuclear Regulatory Commission Washington DC 20555 l
Attention:
Brian K. Grimes, Acting Director i
Division of Reactor Program hianagement Office of Nuclear Reactor Regulation
Subject:
Changes to Design Documentation for the Advanced Boiling Water Reactor (ABWR)
Reference:
- 1. Letter from Brian K. Grimes (NRC) toJoseph Quirk (GE),
" Review of Changes to Design Documentation for the Advanced Boiling Water Reactor (ABWR)," dated April 25,1996.
- 2. Letter (MFN 039-96) fromJ. F. Quirk (GE) to Document Control Desk (NRC), " Submittal of Amendment 36, revision 8, to GE's SSAR and Certified Design Material, Revision 7," dated hf arch 22,1996.
- 3. Letter (MFN 050-96) fromjoseph F. Quirk (GE) to Dennis M.
Crutchfield (NRC), dated April 16,1996.
- 4. Letter from Thomas H. Boyce (NRC) to GE Nuclear Energy,
" Summary of Meeting Held on May 1,1966 to Discuss Changes to the Design Control Document (DCD) for the GE ABWR", Dated May 8,1996.
- 5. Letter (MFN 007-95) fromJack N. Fox (GE) to Document Control Desk (NRC), " Closure of ABWR FSER Confirmatory Item F1.2.2-2" datedJanuary 26,1995. to your letter (Reference 1) provided comments from two NRR organizations on the CDM and SSAR change pages submitted by GE (Reference 2) for alignment with the DCD. Attachment 1 includes markups of DCD and SSAR pages in accordance with comments from TQMB.
9607190012 960610 PDR ADOCK 05200001 f
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e In regards to comments by PDST, GE has recently been informed by yoar staff that Section 16.1 (COL License Information) does not need to be changed, and that other PDST comments regarding this issue in Reference 1 may be disregarded. In addition, PDST requested modifying Section 5A of the Technical Specifications and the DCD Introduction to reflect conformance with the proposed final design certification rule.
GE believes it is appropriate to defer any revision until the status of these issues is determined for the final rule. provides changes to the DCD and SSAR communicated by the staff that are not related to the change packages of Reference 3. Also included in Attachment 2 are Chapter 19 changes to the DCD and SSAR that were identified after the issuance of the FSER. These PRA changes are primarily due to: (1) a reevaluation of the emergency diesel generator and combustion turbine generator capability; (2) correction to the assumed turbine service water isolation features, and (3) an update of the seismic margins analysis. These changes are being incorporated at this time to j
reconcile the DCD and SSAR with the projected FSER Supplement. None of these related changes result in significant impact on the PRA, nor do they alter the ITAAC.
j GE submitted, by Reference 3, proposed changes to the DCD as a result of the First-of-a-Kind-Engineering (FOAKE) detailed design effort. GE and the NRC stafTmet at the NRC on May 1,1996 to discuss the proposed changes. GE prepared the DCD markups resohing the staff comments resulting from the May 1st meeting.
' to Reference 4 includes these markups. GE has identified additional DCD pages affected by the proposed changes; they are included in Attachment 3 along with relevant changes communicated by the staff.
GE requests that your staff review Attachments 1,2 and 3 for inclusion in the ABWR design documentation. Following staff review, GE will submit the updated DCD pages and the corresponding CDM and SSAR pages consistent with the Reference 3 submittal and the markups included in the minutes of the May 1st meeting and the attachments herewith incorporating stafT review comments.
Your letter also requests a description of corrective actions taken to ensure that other design errors are not present in the ABWR DCD. The ABWR FOAKE program, by definition, develops the detailed engineering for the first time for the US ABWR.
This detailed engineering is carried out under our formal design process which l
incorporates thorough re-examination of the DCD design description in light of FOAKE detailed design development work. We also have the benefit of significant nuclear industry oversight and review through the Advanced Reactor Corporation (ARC) in their effort to ensure that GE incorporates industry experience and Utility Requirements Document (URD) design requirements. An integral part of this process is the reconciliation of the detailed design with the DCD design description.
This reconciliation assures that the DCD is correct and accurate. Such revisions to the DCD necessitated by this activity do not alter our belief that the design QA program l
l
we have in place is sound in content and in implementation. Indeed, the relatively few design modifications should enhance confidence in design QA efforts.
The FOAKE program has identified a number of additional desirable design improvements; however, the implementing design changes are not being made at this time since they qualify for post-certification @50.59-type change treatment (i.e., they do not affect Tier 1 or Tier 2* or result in an unrevicwed safety question). Those changes will be made in accordance with governing procedures as established by the Commission.
Finally, as regards the staff's question whether the response to Confirmatory Item Fl.2.2-2 (Reference 5) needs updating, we think not. The ABWR Certification Program hfaster Parts List (hiPL) defines the design documentation implementing the NRC-approved design. The hiPL includes a complete list of all ABWR Certification documents including the DCD and captures the Common Engineering Documents (CEDs) and the Design Action List (DAL). All changes identified by References 3 and 4 and the attachments to this letter will be included in the DCD as i
Revision 3. These changes are being processed for inclusion in Revision 3 by internal GE Engineering Procedures.
Sincerely, k
s J. F. Quirk Project hianager ABWR Certification cc: (w/o attachments)
WT Russell (NRC) l FJ Miraglia (NRC)
TH Boyce (NRC)
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ATTACHMENT 1 SSAR Markups Responding to TQMB Comments 4
I
_ _. - - ~ = - ~
f 23A6100 R:v. 8 ABWR stendentseneryAnetysis nopers k.,
.l in the main stean11ines dovistream of the SRVs. Downstream indications of SRV operation could be changes in such parameters as turbine valve positions or generator output. Osch changes will also be evaluated for anomalies which may indicate a restrictio1 or blockage in a particular SRV tailpipe by making valve-to-valve comparisens. Additionally, during applicaF plant transient testing, where SRVs at: expected to open, operability, operating setpoints, and test pressures will be verified as part of those tests.
(4) Criteria Ievel 1 l
There shall be a positive indication of steam discharge during the manual actuation of each SRV.
Ievel 2 During opening and closing of each SRV, the responses of pressure control system related variables shall be at least quarter-damped (i.e., the decay ratio of the second-to-first overshoot for each variable is less than or equal to 0.25).
The temperature measured by thermocouples on the discharge side of the safety / relief valves shall return to the temperature recorded before the valve l
was opened within 5.6*C range as specified in the GE Startup Test i
Specifications.
During the manual actuation of each SRV, the steam flow discharge through the valve (as measured by change in MWe, BPV position etc.) shall not differ from the average of all the valve responses by more than the limit as specified in the GE Startup Test Specifications.
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14.2.12.2.28 Loss of Feedwater Hosting j
(1) Purpose To demonstrate proper integrated plant response to a loss of feedwater heating event and to verify the ad quacy of the modeling and associated assumptions used for this transient in the plant licensing analysis.
(2) Prerequisites The preoperational tests have been completed and plant management has reviewed the test procedure and approved the initiation of testing. The plant shall be in the appropriate operational configuration with the specified aboMlou 14.2 171 Specific Information to be Includedin Final Safety Analysis Repons - Amendment 36
1 23A6100 Ikv. 3 ABWR standardsafety Analysis Report
(.
14.3 Certified Design Material This section of the SSAR provides the selection criteria and processes used to develop 1
the ABWR Certified Design Material (CDM) that is presented in the GE document 25A5447,"ABWR Certified Design Material." This document provides the principal l
design bases and design characteristics that are certified by the 10 CFR Part 52 rulemaking process and included in the formal certification Rule.
This toplevel design information in the CDM is extracted directly from the more tailed ABWR design information presented in the SSAR (which is part of the rtification application). Limiting the certified design contents to top-level nfortnation reflects the tiered approach to design certification endorsed by the Commission (Staff Review Memorandum 2/15/1991 regarding SECY-90-377; 10 CFR Part 52 Statement of Considerations 54 Fed. Reg.15372,154377, (1989). See also SECY-90-241,90 377 and SECY-91-178.)
The objective of this SSAR section is to denne the bases and methods that were used to develop the CDM document for the ABWR. This SSAR section contains no new l
technical information regarding the ABWR design.
(
The ABWR CDM consists of the following:
(1) An introduction section which defines terms used in the CDM as well as listing general provisions that are applicable to all CDM entries. The intent of these entries is to avoid ambiguities and misinterpretations by providing front-end guidance to users of the CDM.
(2) Design descriptions for: a) systems that are fully within the scope of the ABWR design certification, and b) the in-scope portion of those systems that are only partially within the scope of the ABWR design certification. The intent of the CDM design descriptions is to delineate the principal design bases and i
principal design characteristics that are referenced in the design certification Rule. The design descriptions are accompanied by the inspections, tests, analyses and acceptance criteria (ITAAC) required by 10 CFR 52.47(a) (1)
(vi) to be part of the design certifice. tion application. The ITAAC define verification activities that are to be performed for a facility with the objective of confirming that the plant is built and will operate in accordance with the design certification. Successful completion of these certified design ITAAC, together with the combined license (COL) applicant's ITAAC for the site-specific portions of the plant, will be the basis for NRC("^^^
fuci pcr i; prctic a cf 10 CFR Part 52.103j
%c pea,.y % w (m
i M.3 1 Certified Design Materist - Amendment 33 1
ATTACHMENT 2 DCD and SSAR Markups per Additional Staff Comments (Not related to change packages)
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l R:v.o ABWR Desien coeotcocamentmer2
(4) The flow resistance of open MSIVs is considered. A consenstive value of 2.062 for pressure loss coefficient for two open MSIVs was taken. The nominal value is approximately 3.0. When the open MSIV resistance is considered, the flow chokes at the MSIV on the piping side as soon as the inventory depletion period ends. The effective flow area on the piping side reduces to 70% of a fnctionless piping area. The value of 70% applies to flow of steam and two-phase mixture with greater than 15% quality.
This assumption is quite consenstive because all other resistances in piping are ignored and the flowin the steamline within a one to two second perimiis either all steam or a two-phase mixture of much greater than 15% qu21ity.
os consovvakw e.
(5) MSIVs are completely closed at closing time of 5.5 secondg in order to maximize the]reak flow.av A *
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6.2.1.1.3.3.2.1 Assumptions for Short-Term Response Analysis The response of the reactor coolant system and the containment system during the j
short-term blowdown period of the MSLB accident is analyzed using the assumptions listed in the above subsection and Subsection 6.2.1.1.3.3.1.1 for the feedwater line l
break, with the following exceptions:
(1) The vessel depressurization flow rates are calculated using the Moody's HEM for the critical break flow.
(2) The turbine stop valve closes at 0.2 second. This determines how much steam l
flows out of the RPV, but does not affect the inventory depletion time on the j
piping side.
(3) The break flow is saturated steam if the RPV collapsed water level is below the MSL elevation; otherwise, the flow quality is the vessel average quality. This case provides the limiting drywell temperature.
l Another case was evaluated with the assumption that the two phase level swell would reach the main steam nozzle in one second, thereby changing the flow l
quality to the RPV average quality after one second. This case provides a higher drywell pressure but a lower drywell temperature than the first assumphon.
I l
(4) The feedwater mass flow rate for a MSL break was assumed to be 130% NBR for 120 seconds. This is a standard MSL break containment analysis assumption based on a conservative estimate of the total available feedwater inventory snd the maximum flow available from the feedwater pumps with I
6.2 12 containment Systems
Rev.1 ABWR oesten ceaeosoeceareesmer2 4 (OpaA ssASma.,lp Table 14.3-3 Transient Analysis Tier 2 Entry Parameter Tier 2 Value Table 15.0-1 Reactor internal Recirculation Pumps Number of Pumps 10 Pump Trip Inertia (kg m2)
Trip Mitigation-(maximum) 26.B Accident (minimum) 17.5 Relief Valve (Relief Function) l Capacity (% NBR Steam Flow at 7.89 MPaG) 91.3 Number of Valves 18 Opening Time (s) 0.15 i
(valve stroke time only. Does not include 0.1 s delay to energize solenoid)
High Flux Trip Scram APRM Simulated Thermal Power Trip Scram Total Steamline Volume (m3) 113.2 l
Table 15.0-6 FMCRD Scram Times 10% Rod insertion (s) 0.46 40% Rod insertion (s) 1.208 60% Rod insertion (s) 1.727 100% P.od insertion (s) 3.719 15.1.1.2.2 High Simulated Thermal Power Trip Scram l
Table 15.1-5 High Water Level 8 Initiates Feedwater Pump Trip Table 15.1-5 Turbine Stop Valve Position Switches initiate l
Reactor Scram l
Trip of 4 RIPS Table 15.1-7 Low Water Level 2 initiates Trip of 6 RIPS RCIC System l
Maximum Startup Time ((s)- includes 1.0 s for 30 instrument delay) l MSIV Closure on Low Turbine inlet Pressure
)
i 15.1.3.3.1
".'.s..r. MSIV Closure Time (s esewmos 0.5 s for 5.0 1
l Instrument delay) l Womwm ss o k h o n v o\\v e, clo s m j
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l 14.3-28 Tier 1 Selection Criteria and Processes
DCD Rev.1 ABWR oasion entroloocamnoner2
+ (O f da4< SSM L%Wl) 3 Table 14.3-3 Transient Analysis (Continued)
Tier 2 Entry D3 Parameter Tier 2 Value Table 15.1-9 SRNM gh Nytr in Flux Scram 15.1 1.3.1 TCV Full oke rvo Closure (s) 2.5 Table 15.2-1a Low Water Level 3 Initiates Trip of 4 RIPS Table 15.2-2 High Dome Pressure initiates Trip of 4 RIPS Table 15.2-3 T/G Load Rejection initiates Turbine Ccntrol Valve Fast Closure Turbine Bypass System Operation on High Pressure Fast Control Valve Closure Initiates Scram Trip of 4 RIPS 15.2.2.3.1 TCV Full Stroke Fast Closure (s - from normal operating position) 0.08 Table 15.2-6 Turbine Trip initiates Turbine Control Valve Fast Closure Turbine Bypass System Operation on High Pressure 15.2.3.3.1 Turbine Stop Valve Full Stroke Closure (s) 0.10 Table 15.2-9 MSIV Position Switches initiate Scram 15.2.4.3.1 Minimum MSIV Closure Time (s) 3.0 Table 15.214 Low Condenser Vacuum initiates MSIV Closure 15.2.6.1.1.2 RIP M/G Set Number of RIPS 3
Length of Time Hold Original Speed (s)
0 RIP Coastdown Rate (% per s) 10 Length of Time (s) 2.0 Time of RIP Trip (s) 3.0 Table 15.2-17 Low Water Level 3 Initiates Reactor Scram l
15.2.7.2.21 Meets Single-failure Criterion 15.2.9 RHR System has 3 Independent Divisions 15.3.1.1.1 No More Than 3 RIPS on One Electrical Power Bus 15.3.1.2.2.2 Rapid Core Flow Coastdown initiates Reactor Scram j
M 3 29 l
Tier 1 Selection Criteria and Processes
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ABWR oesion controloccamentmer2
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expected during the course of the event. Once isolation eccurs, the pressure will increase to a point where the SRVs open. The operator should:
(1) Monitor that all rods are in l
l (2) Monitor reactor water level and pressure l
(3) Observe turbine coastdown and break vacuum before the loss of steam seals.
l Check turbine auxiliaries i
(4) Observe that the reactor pressure relief valves open at their setpoint l
(5) Observe that RCIC initiated on low-water level (6) Secure RCIC when reactor pressure and level are under control (7) Monitor reactor water level and continue cooldown per the normal procedure (8) Complete the scram report and initiate a maintenance survey of the SB&PCS l
before reactor restart 15.1.3.2.2 Systems Operation 15.1.3.2.2.1 Inadvertent Opening of One Turbine Bypass Valve This event does not require any protection system or safeguard system operation. This analysis assumes normal functioning of plant instrumentation and controls.
15.1.3.2.2.2 Inadvertent Opening of All Turbine Control Valves and Bypass Valves l
To properly simulate the expected sequence of events, the analysis of this event assumes l
normal functioning of plant instrumentation and controls, plant protection and reactor l
protection systems, except as otherwise noted.
Initiation of RCIC System functions occurs when the vessel water level reaches the L2 setpoint. Normal startup and actuation can take up to 30 seconds before effects are realized.
l If these events occur, they will follow sometime after the primary concerns of fuel thermal margin and overpressure effects have occurTed, and are expected to be less severe than those already experienced by the system.
15.1.3.3 Core and System Performance 15.1.3.3.1 input Parameters and initial Conditions A five-second isolation valve closurejnstead of a three second clos
( w, w n,s.I d,o, a s lu e l o >..9 h~t P % m)
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the turbine pressure decreases below the turbine inlet low pressure setpoint for main Decrease in Reactor Coolant Temperature 15.1 9
Rev.O ABWR oesie coneetoecamenttrierz
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%s seeb^ca W osda.nJo30 " U of a p(bowa ecM bM*
15.2.4.3 Core and System Performance J4ev G1bd T.B*y* Md 15.2.4.3.1 input Parameters and initial Conditions ow s/23 9C 1
g The main steam isolation valves close in 3 to k secondsyThe worgt case (the 3 second closure time) '
med in this analysis. No e red d wa.s +awew for s erba d e.\\og.
Position switches on the valves initiate a reactorscram when the valves are less than 85%
open. Closure of these valves causes the dome pressure to increase. Four RIPS are tripped when the high pressore setpoint is reached.
ABWR has motor-driven feedwater pumps. However, a conservative feedwater flow coastdown model was used in order to bound both the motor-driven and steam turbine driven feedwater pump designs.
15.2.4.3.2 Results 15.2.4.3.2.1 Closure of All Main Steamline isolation Vdves Figure 15.2-9 shows the changes in important nuclear system variations for the simultaneous isolation of all main steamlines while the reactor is operating at 102% of NBR power. Neutron flux increases slightly, and fuel surface heat flux shows no increase Four RIPS are tripped due to high pressure. Water level decreases sufficiently to cause a trip of remaining 6 RIPS and the initiation of the RCIC system on the Level 2 (L2) trip at some time greater than 10 seconds. However, there is a delay up to 30 seconds before the water supply enters the vessel. Nevertheless, there is no change in the thermal margins. Therefore, this event does not have to be reanalyzed for specific core configurations.
15.2.4.3.2.2 Closure of One Main Steamline isolation Valve Only one isolation valve is permitted to be closed at a time for testing purposes to prevent scram. Normal test procedure requires an initial power reduction to approximately 75 to 80% of design conditions in order to avoid high flux scram, high pressure scram, or full isolation from high steam flow in the " live" lines. With a 3 second closure of one MSIV during 102% rated power conditions, the steam flow disturbance may raise vessel pressure and reactor power enough to initiate a high neutron flux scram. This transient is considerably milder thar. closure of all MSIVs at full power. No quantitative analysis is furnished for this event. However, no significant change in thermal margins is experienced and no fuel damage occurs. Peak pressure remains below SRV setpoints. Therefore, this event does not have to be reanalyzed for specific core configurations.
l Increase in Reactor Pressure 15.2 15 1
r l
l Rev. 1 ABWR oasion contrescoenmeatmer2
>- (.Upaa s3 AA Sw aj) break. This level of activity is consistent with an offgas release rate of 3.7 GBq/s for Case 1 and 14.8 GBq/s for Case 2 referenced to a 30 minute decay.The iodine concentration in the reactor coolant is:
MBq/g Case 1 Case 2 I-131 0.001739 0.03515 I-132 0.01536 0.30747 I-133 0.01206 0.24161 l
I-134 0.02634 0.52688 l
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I-135 0.01647 0.3293 Other isotopes of high intrinsic activity such as N-16 have been precluded due to their extremely short halflives.
I 15.6.4.5.1.2 Fission Product Transport to the Environment o
l The transport pathway is a direc nfiltered release to the environment. The MSIV l
detection and closure time of 5. sfesults in a discharge of 12,870 kg of steam and l
21,953 kg ofliquid from the break. Assuming all the activity in this discharge becomes l
airbohrelease of activity to the emironment is oresented in Table 15.6-6
(( r u, - - fi s t v e.l s y 7. N e, a no3 tr%ea,(ety}
l 15.6.4.5.1.3 Results l
The calculated exposures for the design basis analysis are presented in Table 15.6 7 and are less than the guidelines of10CFR100. COL applicants need to update the calculations to conform to the as-designed plant and site-specific parameters (see Subsection 15.6.7.2 for COL license information.).
15.6.5 Loss-of-Coolant Accident (Resulting from Spectrum of Postulated Piping t
i Breaks Within the Reactor Coolant Pressure Boundary)-inside Containment This event postulates a spectmm of piping breaks inside containment nrying in size, type, and location. The break type includes steam and/or liquid process system lines.
This event is also assumed to be coincident with a safe shutdown earthquake (SSE) for the mechanical design of components.
The event has been analyzed quantitatively in Sections 6.3 (Emergency Core Cooling Systems),6.2 (Containment Systems),7.3 and 7.1 (Instrumentation and Controls), and Decrease in Reactor Coolant inventory 15.6 4
Rev.1 ABWR oesion controsoocemeatmerz S.wloA )
(Opdate ss A-A.
J Table 15.6-3 Instrument Line Break Accident Results Meteorology
- and Dose Results Meteorology Distance Thyroid Dose Whole Body Dose 3
(s/m )
(m)
(Sv)
(Sv)
=
8.59E-03 max 3.0E-01 6.0E-03 1.37E-03 Chp 2 4.8E-02 9.4E-04 2.19E-04 800 7.6E-03 1.5E-04 i
1.11 E-04 1600 3.9E-03 7.9E-05 5.61E-05 3200 2.0E-03 4.0E-05 3.73E-05 4800 1.3E-03 2.6E-05 l
- Meteorology calculated using Regulatory Guide 1.145 for a ground level 1.0 m/s, F stability release.
" Max" = maximum meteorology to meet 10% of 10CFR100 limits.
Table 15.6-4 Sequence of Events for Steamline Break Outside Containment Time (s)
Event 0
Guillotine break of one main steamline outside primary containment.
-0.5 High steamline flow signalinitiates closure of main steamline isolation valve
<1.0 Reactor begins scram.
l y_.5df 6.0 Main steamline isolation valves fully closed.
38 Safety / relief valves open on high vessel pressure. The valves open and close to maintain vessel pressure at approximately 7.58 MPa.
30 RCIC initiates on vessel low-water Level 2.
50 RCIC begins injection.
l l
199 HPCF initiates on low water level.
236 One HPCF begins injection (the other HPCF is unavailable due to the single failure assumption).
1-2 hours Normal reactor cooldown procedure established.
l l
l 1 646 Decrease in Reactor Coolant Inventory
R:v. 2 ABWR onie cutroioocmutwrier2 4 (.Updnic ssMLF.m.Iung Table 15.6-5 Steamline Break Accident Parameters i
Data and assumptions used to estimate source terms.
A. Power Level 4005 MWt B. Fuel damage none C. Reactor coolant activity Subsection 15.6.4.5 D. Steam mass released 12,870 kg E. Water mass released 21,953 kg 11 Data and assumptions used to estimate activity released A. MSIV closure time (break time Ks
- 5. D l
until fully closed)
- 8. Maximum release time 2h 111 Dispersion and Dose Data A. Meteorology Table 15.6-7 B. Boundary and LPZ distances Table 15.6-7 C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions Reference 15.6-2, RG 1.109, and ICRP 30 E. Activity Inventory / release Table 15.6-6 F.
Dose Evaluations Table 15.6 7 Decrease in Reactor Coolant Inventory 15.6-27 1
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19P Evaluation of PotentialModificationsto the ABWR esign This Section is not part o the DCD (Refer to Attachment A of the Technical Support Document for the ABh
). Attachment A of the Technical Support Document for the ABWR is not incorporated by reference in the DCD.
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19P Evaluation of Potential, Modifications to the ABWR esign i
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19P V2 Evatustion of Potential Modifications to the ABNR Design-Amendment 36
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ABWR 4
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An ATWS condition exists when either of the following occurs:
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-(a) High RPV pressure (7.76 MPaG) and startup range neutron monitor 0
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(SRNM) not downscale for 3 minutes, or SRNM (b) Low RPVlevel (Level 2) and not downscale for 3 minutes.
A light in the control room indicates that power is avadable to the pump motor contactor and that the contactor is deenergized (pump not running). Another light indicates that the contactor is energized (pump running).
Storage tank liquid level, tank outlet valve position, pump discharge pressure, and l injection valve position indicate that the system is functioning. If any of these items l indicates that the liquid may not be flowing, the opemtor shall immediately change the other switch to the START mode, thereby acovanng the redundant train of the SLCS.
The local switch cannot prevent the operation of the pump from the control room.
Pump discharge pressure and valve status are indicated in the control room.
Equipment drams and tank overflow are not piped to the Radwaste System but to r
separate containers (such as 208L drums) that can be removed and disposed o independently to prevent any trace of boron from inadvertently reachmg the reactor 1
Instrumentation consisting of solution temperature indication and control, solution level and heater system status is provided locally at the storage tank. Table 9.3-1 co the process data for the various modes of operation of the SLCS. Seismic cat quality class are included in Table 3.2-1. Principals of system testing are disc' Subsection 9.3.5.4.
f 9.3.5.3 Safety Evaluation The SLCS is a reactivity control system and is mamtainedin an operable status whenever the reactor is critical. The system is never expected to be needed for safety reasons because of the large number ofindependent control rods avadable to shut down the reactor.
To assure the availability of the SLCS, two sets of the components required to actuate the system (pumps and injection valves) are provided in parallel redundancy.
The system is designed to bring the reactor from rated power to a cold shutdown a time in core life. The reactivity compensation provided will reduce reactor power from rated to zero level and allow cooling of the nuclear system to room temperature, with l
the control rods remaining withdrawn in the rated power pattem. It includes the reactivity gains that result from complete decay of the rated power xenon inve also includes the positive reactivity effecu from eliminating steam voids, changing wa density from hot to cold, reduced Doppler effectin uranium, reducing neutron l from boiling to cold, and decreasing control rod worth as the moderator cools.
s.w Process Auxiliaries
23A6700 kv. 8
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ABWR swwardsarery Analysis Report
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An A'RVS condition exists when either of the following occurs:
(a) High RPV pressure (7.76 MPaG) and startup range ron monitor
(@t) not downscale for 3 minutes, or sMM (b) Low RPV level (Level 2) and not downscale for 3 minutes.
A light in the control room indicates that power is available to the pump motor contactor and that the contactor is deenergized (pump not running). Another light indicates that the contactoris energized (pump running).
Storage tank liquid level, tank oudet valve posidon, pump discharge pressure, and injection valve position indicate that the system is funcdoning. If any of these items indicates that the liquid may not be flowing, the operator shall immediately change the other switch to the START mode, thereby activadng the redundant train of the SLCS.
The local switch cannot prevent the operation of the pump from the control room.
Pump discharge pressure and valve status are indicated in the control room.
Equipment drains and tank overflow are not piped to the Radwaste System but to separate containers (such as 208L drums) that can be removed and disposed of l
independendy to prevent any trace of boron from inadvertently reaching the reactot.
Instrumentation consisting of solution temperature indicadon and control, solution I
level and heater system status is provided locally at the storage tank. Table 9.31 contains the process data for the various modes of operadon of the SLCS. Seismic category and quality class are included in Table 3.2-1. Principals of system testing are discussed in Subsecdon 9.3.5.4.
j 9.3.5.3 Safety Evaluation The SLCS is a reactivity control system and is maintained in an operable status whenever l
the reactor is cridcal. The system is never expected to be needed for safety reasons l
because of the large number ofindependent control rods available to shut down the reactor.
To assure the availability of the SLCS, two sets of the components required to actuate l
the system (pumps and injecdon valves) are provided in parallel redundancy.
I The system is designed to bring the reactor from rated power to a cold shutdown at any l
time in core life.The reacti ity compensation provided will reduce reactor power from rated to zero level and allow cooling of the nuclear system to room temperature, with the control rods remaining withdrawn in the rated power pattern. It includes die j
reactivity gains that result from complete decay of the rated power xenon inventory. It also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced Doppler effectin uranium, reducing neutron leakage from boiling to cold, and decreasing control rod worth as the moderator cools.
9.3 13 ho:ess Auxilianes - Amendment 36
DCD CHAPTER 19 CHANGES (Except Section 19P)
l l
i R>v. 0 4
ACWR oesian contratosarnse:ttri:o 2 4
i i
i Table 1.9-1 Summary of ABWR Standard Plant COL License information (Continued) t i
item No.
Subject Subsection 19.8 Action to Avoid Common-Cause Failures in the Essential 19.9.8 i
Multiplexing System (EMUX) and Other Common-Cause Failures i
19.9 Action to Mitigate Station Blackout Events 19.9.9 19.10 Actions to Reduce Risk of Intemal Flooding 19.9.10 t'
19.11 Actions to Avoid Loss of Decay Heat Removal and 19.9.11 Minimize Shutdown Risk 19.12 Procedures for Operation of RCIC from Outsde the 19.9.12 Control Room 19.13 ECCS Test and Surveillance intervals 19.9.13 19.14 Accident Management 19.9.14 19.15 Manual Operation of MOVs 19.9.15 19.16 High Pressure Core Flooder Discharge Valve 19.9.16 19.17 C' ability of Containment isolation Valves 19.9.17 19.18 Procedures to Ensure Sample Lines and Drywell Purge 19.9.18 Lines Remain Closed During Operation 19.19 Procedures for Combustion Turbine Generato 19.9.19 QEmeroencv Diesel Generatof to Supply Power to Condensate Pumps 19.19a Actions to Assure Reliability of the Supporting RCW and 19.9.20 Service Water Systems 19.19b Housing of AICWA Equipment 19.9.21 19.19c Procedures to Assure SRV Operability During Station 19.9.22 Blackout 19.19d Procedures for Ensuring Integrity of Freeze Seals 19.923 19.19e Procedures for Controlling Combustibles During 19.9.24 Shutdown 19.19f Outage Planning and Control 19.9.25 19.19g Reactor Service Water Systems Definition 19.9.26 19.19h Capability y uu r kers 19.9.27 19.191 Capability o Con '
ent Atmosphere Monitoring 19.9.28 System 19.19]
Plant Specific Safety-Related issues and Vendors 19.9.29 Operating Guidance 19.20 Long-Term Training Upgrade 19A.3.1 1.9-12 COL LicenseInformation
.e i l
ctv. o ABWR Desina cueelcunnavrie2 Table of Contents (Continued)
(
19'.6.9 Safe ty Goal Policy Statem en t..........................................................
.19.6-6 19.6.10 NotUsed............................................................................................19.66 19.6.I1 Conclusion..........................................................................
.19.6-6 19.6.12 Re fe r e n c e s........................................................................................... 19.6 6 19.7 PRA as a D esign Tool........................................................................................... 19.7-1 19.7.1 ABWR Design and Operating Experience.................................................... 19.7-1 19.7.2 Early PRA Studies......
.............................................................................19.7-1 19.7.3 PRA Studies During the CertiScation Effort................................................ 19.7-3 19.7.4 Con duct of the PRA Evaluations................................................................ 19.7-10 19.7.5 Evaluation of Potential Design Improvements.......................................... 19.7-10 19.8 Important Features Identified by the ABWR PRA...................................................... 19.81 19.8.1 Important Features from Level 1 Internal Events Analyses.........................19.8-2 19.8.2 Important Features from Seismic Analyses.................................................... 19.8 8 19.8.3 Important Features from Fire Analyses......................................................... 19.810 19.8.4 Important Features from Suppression Pool Bypass and Ex-Containment LO CA An alyses........................................................................................... 19. 8 1 2 19.8.5 Important Features from Flooding Analyses............................................ 19.816 19.8.6 Important Features from Shutdown Events Analyses.................................19.819 19.8.7 ABWR Features to Mitigate Severe Accidenu............................................... 19.8-22 19.9 CO L Lic ense Info rm atio n.................................................................................... 19.9 1
{..
19.9.1 Post Accident Recovery Procedure for Unisolated CUW Line Break..........
19.9-1 19.9.2 Confirmation of CUW Operation Beyond Design Bases............................19.9 2 19.9.3 Event Specific Procedures for Severe External Flooding....................... 19.9 2 19.9.4 Confinnation of Seismic Capacities Beyond the Plant Design Bases............19.9-3 19.9.5 Plan t Walkdowns........................................................................................... 19.9 3 19.9.6 Confirmation of Loss of AC Power Event................................................... 19.9-4 19.9.7 Procedures and Training for Use of AC. Independent Water Addition j
System..............................................................................................................19.9-4 19.9.8 Actions to Avoid Common.Cause Failures in the Essential Multiplexing System (EMUX) and Other Common.Cause Failures................................ 19.9-5 19.9.9 Actions to Mitigate Station Blackout Events........................................... 19.9-5 19.9.10 Actions to Reduce Risk of Internal Flooding...........................................
19.9-6 19.9.11 Actions to Avoid Loss of Decay Heat Removal and Minimize Shutdown Risk........................................................................................................19.9-7 19.9.12 Procedures for Operation of RCIC from Outside the Control Room........19.9-9 19.9.13 ECCS Test and Surveillance Intervals................................................... 19.9-10 19.9.14 Ac cid e n t M an agem e n t....................................................................... 19.9-10 19.9.15 Manual Operation of M OVs....................................................................... 19.9-1 1 l
19.9.16 High Pressure Core Flooder Discharge Valve...................................... 19.9-11 19.9.17 Capability of Containment Isolation Valves........................................ 19.9-11 19.9.18 Procedures to Insure Sample Lines and Drywell Purge Lines Remain Closed During Operation.....................................................)2.,,,,, i 9,9.i 3 19.9.19 Procedures for Combustion Turbine Generator End i meroency DieseD encrato to Supply Power to Condensate Pumps.............
.... 19.9-11 19.ii Table of Contents
.. ~....
R:v. O ABWR oneen conedoecamantra2
\\
Considedng these composite bounding scenarios is an added conservatism to the already conservative FIVE methodology.
Fire ignition frequencies were developed for each of the above scenarios by directly applying the prescdptive steps documented in the FIVE methodology. Bounding core damage frequency estimates were developed by applying these initiating event frequencies to appropriately modi 6ed ABWR Level 1 fault and event tree models and reevaluating them.
The final bounding core damage frequency for each of the Sve scenarios was calculated and determined to be acceptable. These results reflect the inherently conservative nature of the FIVE methodologyitself, compounded byits additional conservanve application in evaluating fire impact at the dmsional fire area, control room complex, and turbine building fire levels. Addressing ABWR fire risk at the fire compartment level, considering ignition <anree< fire oro.zrdan =nd =nanre<<ian in more dermiN" redner thicvalue fsleaks A b Me 34sw New nse ws iMuewd W 'EhGEC M YI'[.Y $'
Ah YL N
19.4.5 Anwe r os..;;;.a A
- ..g analys s The results of the ABWR Probabilistic Mooding Analysis show t the turbine, control, and reactor buildings are the only structures that req uations for potential
,p j
flooding. The other buildings do not contain any equip ent that could be used for safe shutdown or potential flooding would not result in lant transient.
9[}-0o Flooding in the turbine building could result in turbine trip due to loss of circulating 5g water or feedwater. Automatic pump trips an e closure on high waterlevel should terminate the flooding. Butif these were to
, a non-watertight door at grade level in
$p,
ud 4 the turbine building should allow water
't the building. If this door retained water, 6
watertight doon would prevent water e ring the control and reactor buildings. The oO) core damage frequency (CDF) for e building floodmg is extremely small.
The worst case flood in the control b ' ding is a break in the reactor service water system 1
-5 Q (RSW) which is an unlimited sourc. Floor drams and other openings in the floorwould
$o direct all flood water to the first r where the reactor component cooling water t$
(RCCW) rooms are located. Th CCW rooms contain sump pumps. Water level sensorsin the RCCW rooms sh d actuate alarms in the control room and send signals
?7 N h3 to tdp the RSW pumps and cl isolation valves in the RSW system. If these sensors Ej were to fail, watertight doo on each room should limit flood damage to only one of o
the three RCCW divisions.
e CDF for control building flooding is extremely small.
Reactor building flooding could occur either inside or outside secondary containment.
l In either case, the flooding sources are finite with the suppression pool and condensate storage tank being the largest sources. Inside secondary containment flooding cannot cause damage to equipment in more than one of the three safety divisions because of 3
oA Sin + 4toor sp.12 External Event Analysis and Shutdown Risk Analysis
Rev. O come coentoecameetmea ABWR Je
%n k.a.,Vs w \\ sus ad ap 9%avs wa t Ir<.,
asW b 4\\ow ch aw s l,,,,, '
3 k.
watertight doors on each safety division room. As was the in the control building,
^ ^ + 2x r.dc,i r 7 ' ; "" h -~ C,J. -.
sump pumps on the first floor. The available volume of rooms on the first floor can contain all potential flood sources. Outside secondary contamment, floor drains direct all flood water to sump pumps on floor B1F (third floor). If the sump pumps fail or cannot keep up with the floodmg rate, an overfillline in the sumps direct water to the corridor of the first floor where it can be contained as diem==ad above. The CDF for reactor building flooding is extremely small.
Lbe b uunal Alood$g 9 The total CDF for internal flooding is very small.
Otcwe bd Noor ales m \\\\
li %,4 4.k. w be da.%b 19.4.6 ABWR Shutdown R sw e L. +k + w. do.
% s.fd '
Ma A3 wqwun.d mil o c.
The ABWR ign has been evaluated for risks with shutdown conditions (i.e.,
ode 3, 4 and 5). The evaluation included the following shutdown risk categories:
J (1) Decay heat removal, (2) Inventory control, k
(3). Containmentintegrity, (4) Loss of electrical power, (5) Reactivity control.
The evaluation also included risk reduction features of the ABWR due to j
instrumentation, flooding and fire protection, use of freeze seals, and procedure guidelines. ABWR features that are not part of current BWR designs were evaluated to determine if any new vulnerabilities would 'oe introduced. In addition, an evaluation of approxunately 200 events at operating BWR plants which were considered precursors to loss of decay heat removal capability showed that ABWR design features could mitigate the effects of all these events.
i l
The results of the ABWR shitdown risk analysis demonstrated that the core damage j
frequency (CDF) for all shutdown eventis very small._he main features that contribute T
to this low CDF are:
4 (1) Three physically and electrically independent residual heat removal (RHR)
}
and support systems.
(2) Multiple makeup sources for inventory control (e.g., suppression pool, l
condensate storage tank, AC independent water addition system).
4 I M -13 Extemal Evern Analysis and Shutdown Misk kalysis J
-l i
Rev.0
.ABWR oess n comretoocumeownerz o
I The capability of the Reactor Water Cleanup (CUW) System to provide an additional f
means of decay heat removal with the reactor at high pressure wasjudged to be less important than the features selected as "important features." The additional
=
redundancy provided by this capability does not significantly reduce the calculated ABWR CDF. This is due to the high reliability of other means of decay heat removal such i
as the various modes of operation of the three RHR loops and the containment overpressure protection system which result in a very small contribution of Class II sequences to total CDF without the CUW capability.
The degree of redundancy in SRVs to perform the ADS function was alsojudged to be i
less important than other features. Only three SRVs are required to open to depressurize the reactor so that low pressure pumps can provide the necessary cooling.
The eight ADS SRVs plus the remaining ten SRVs that can be manually actuated far exceed redundancy requirements for depressurization. ADS failure is dominated by common cause failure of the ADS valves.
Another featurejudged to be less important than other features is the automatic l
initiation of RHR on suppression pool high temperature. Many hours are available to initiate RHR to remove heat from the suppression pool following transients that dump heat to the suppression pool. The reliability of operators to manually initiate this i
function when required isjudged to be very high, therefore this automatic initiation feature does not significantly reduce the cal.ulated CDF.
The capability to manually initiate scram wasjudged to be less important than the selected features. The ability to manually initiate scrum is not an important feature from the standpoint of CDF due to the highly reliable, redundant, and divene features of the reactivity controlsystems.
y The capability to use the CRD hydraulic system to provide additional water injection into the core wasjudged to be less important than the selected features. This is the primary reason that numerical credit for this function was not taken. The valve in some sequences for the coolant injection capability of the CRD pumps is of a lesser importance since adequate core cooling is available from other sources to assure a very low core damage frequency.
It was alsojudged that the high drywell pressure signal for ADS was less important than the selected features. With the incorporation of the drywell high pressure signal bypass timer, the high drywell pressure signal for ADS is less important.
m The capability to provide AC power to the condensate pumps from a safety-rela.ed bus was oncejudged to be an important feature. However, its importance decreased after j
l the combustion turbine generator was added to the design. If no credit were taken for condensate pumps being powered by safety buses, core damage frequency increases by a small percentage.
Important Features Identdied by the ABWR PRA 19.8 7 r
l
Rsv.O
\\
ABWR oesira comroscocumntmer2 1
19.9.15 Manual Operation of MOVs As noted in Subsection 19.7.3 (3)(a), manual operation of MOVs can be used to l
improve the availability of decay heat removal. The COL applicant will implement procedures for such an operation.
19.9.16 High Pressure Core Flooder Discharge Valve As noted in Subsection 19D.7.7.5, the HPCF loop B pump discharge valve is in the drywell. Plant procedures should include independent verification that the valve is locked-open following maintenance.
19.9.17 Capability of Containment isolation Valves i
l To insure that containment isolation valve capability does not reduce the containment capability, the COL applicant will demonstrate that the stresses ofcontainment isolation valves, when subjected to severe accident loadings of 0.77 MPa internal pressure and 260'C (500'F) tempemture in combination with dead loads, do not exceed ASME Section III service level C limits. In addition, the ultimate pressure capability at 260*C l
(500*F) will be shown to be at least 1.03 MPa.
l 19.9.18 Procedures to insure Sample Lines and Drywall Purge Lines Remain Closed i
During Operation As noted in Subsection 19.8.4.3, it is important that these lines be normally closed l
during plant operation. The COL applicant will develop procedures and administrative controls to ensure the valves are normally sealed closed and that the purge valves have l
motive power to the valve operators removed.
19.9.19 Pr=dures for Combustion Turbine Generatorfnd Emergency Diese enerat@to Supply Power to Condensate Pumps l
The COL applicant will implement procedures for manual er of Combustion Turbine Generator (CTG) ana t.mergenev niewl Generator (FDGhpower to the l
condensate pumps..t nrrd in %bx;in 10.0.1.S, J., -L L.; cf 9-E"C :: p;= Se
=xi. _ y.;;:p-~ 9 m enr-A" " b..._ _.m
..y.
l i: S p :52
.a m,,
- r. c gg.s. crn eaa
~
- W. Condensate pump support systems (lube oil, cooling water) are also n id if or pur;: r:.. p.n.it
==. w de PW fr a.b.cni' r..ci ">:-1;;q W p% Q.e 4k CTG.
l 19.9.20 Actions to Assure Reliability of the Supporting RCW and Service Water Systems t
To assure the reliability of the RCW and Service Water Systems, the COL applicant will l
take the following action. At least each month, the standby pumps and heat exchangers are started and the previously running RCW and service water equipment is placed in a l
standby mode.
l l
COL License Information 19.9 11 i
- _ ~ -
nov. o ABWR Deelna conedDecamenener2 I
~,
in a high-pressure core melt since all control is lost, the high-pressure systems fail, and the reactor cannot be depressurized. The condition of successful emergency DC and AC power and successful scram is indicated by the ET transfer and is descdbed in detailin Figure 191-2. The condition of successful emergency DC and AC power, but with failure to scram is indicated by the ATWS transfer, and is desenbed in Figure 19I-3.
i The condition of successful emergency DC and failure of emergency AC continues on Figure 191-1. The next question is whether or not there is a failure to scram (node C).
l Failure to scram is considered as a Class IV core melt. With successful scram, RQC (node UR) and firewater (node FA) are the only available means ofwater injection into the RPV since all AC power is lost. Since station batteries will eventually discharge resulting in loss of RCC, or if ROC fails, the reactor must then be depressurized (node X) to allow firewater injection. The loss of emergency DC power (station batteries) results in a high-pressure core melt as shown in Figure 191-1.
The firewater system has a diesel driven pump and all needed valves can be accessed and l
operated manually. No support systems are required for firewster operstion. See Subsection 19.9.21 for COL license information pertaining to housing of AClWA equipment. The random failure probability of firewater is dominated by operator l
failure to initiate the system. For the upper branch, where ROC is successful, the l
operator has 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before the station batteries expire and RQC trips. The human error probability (HEP) for this case is very small. For the lower branch, where ROC fails, the operator has only 30 minutes in which to depressunze the reactor and initiate l
l l
Srewater injection. For this case, the HEP is moderate. In the event that the firewater
-l diesel fails to start, the operator could make use of a fire truck, but this was not modeled.
If the RHR heat exchanger fails (node HX) due to the earthquake, it is presumed that the failure could include a pipe break that could pamally drain the suppression pool into the RHR pump room. These core damage sequences are identified with a "P" (e.g.,
IB2-P). Fission product scrubbing would still be effective in preventing a large release.
The effects of possible flooding on equipment operation beyond the RHR room were considered and found to be relativelyinsignificant because of the relatively high HCLPF of the heat exchangers, the ability of the operator to isolate the break, and the presence of the independent ACIWA (firewater) system.
191.3.2 LOOP with Emergency Power and Scram Event Tree In the event tree of Figure 191-2 (ET transfer), there are two simdar divisions depending j
on whether or not there is a stuck open relief valve (node PC). If there is a stuck-open valve, the reactor will eventually depressurize causing loss of RCC steam supply. The l
probability of having a stuck-open valve is based on operating experience. If both high-pressure injection systems fail, the reactor must be depressurized rapidly for low-g l
l pressure system use (LPFL-V1hr condensate injection -V2). In ABWR, condensate pumps can oe transferred to the emergency bus by the operator.
1914 l
Seismic Margine. J,2 l
My. 0 ABWR omin onedanamensa C
191.3.3 ATWS Event Tree Figure 191-3 (ATWS transfer) represents failure to scram, and requires standby liquid control (automatic) and operator action to control reactor water levelwith the injection system (s) that are available. The HEP for this action is small. In this A'IWS analysis, if high-pressure systems fail, core damage results. No credit is given to low-pressure l
l injection. For an ATWS, the probability of a stuck open SRV was conservatively l
increased on the basis ofincreased SRV activity.
l l
191.4 System Analysis The fault trees used in the seismic systerr alysis are shown on Figures 1914 through 191-15. The seismic system analysis calce.e. ac probability of seismic failure and corresponding system HCLPFs of each ot the important systems throughout the seismic l
ground acceleration spectrum. The system HCLPFs are then input to the event trees and combined with random system failure probabilities and human errors. The seismic fault trees contain only those componenu that might be subject to seismic failure.
Random system failure probabilities are taken from the internal events analysis and include all other components. One of the important ground rules of the seismic margins analysis is that alllike components in a system always fail together.
The reactor protection system, control rod drive system, and alternate rod insertion system were not modeled since the failure of control rods to insert is dominated by the relatively low seismic fragdity of the fuel assemblies, control rod guide tubes, and housings. A seismic fault tree for reactivity control is shown on Figure 191-15. The fuel assemblies are the most fragile component.
A seismic fault tree for the standby liquid control system is shown on Figure 19114.
Failure of the standbyliquid control system is dominated by failure of two components:
the pump and boron supply tank.
Since the most fragile essential component in the plant is the ceramic insulator in the switchyard, the loss of offsite power dominates the analysis and the availability of emergency power becomes very important. The lose.of-power fault tree (Figure 191-10) is for emergency AC power. In the loss of emergency AC power fault tree, the more l
fragile components are the diesel generator, transformers, motor control centers.
inverter and circuit breaker. The DC power fault tree (Figure 19I-11) has two elements:
4 batteries and cable tray.
A m g.4, % 9 *w Systems and equipment which require offsite power, such as the feedwater syste are not modele_d since offsite power is presumed to be not avadable for the core damage f
sequencepse condensate injection system is modeled on Figure 191-15 since credit iA i
en to the operator for transfernng condensate to an emergency bus (Figure 191-2.)
The human error probability is m-b=='ar 'han the seismically induced equipmenM SeismicMarginsAnalysis 19M
Mov. 0 l
ABWR aneine coneetoncemenmea l
k
~
i failure probability, therefore, this fault tree has negligible impact on the HCLPF valu I
of the co:Tesponding accident sequences.
l l
Essential service water is as important as emergency power, and its loss would have much the same effect as the loss of emergency AC power. The loss of4ervice-water fault tree is shown on Figure 191-12. The more fmgile components in this system are the service water pump, heat exchanger, and room air conditioning unit. The service water pump house is also included in this fault tree.
i Structure failures that could contribute to seismic core damage are shown on Figure 191-9. In this analysis, any one or more of these structural failures are conservatively presumed to result in core damage. The structures having the lowest seismic capacity are the reactor building and control building.
The remainder of the fault trees are for core cooling (Figures 191-4 through 1914). The more fragile components in these systems are the pumps, heat exchangers, and the l
firewater supply tank. The condensate storage tank (CST) is not modeled since the ECCS systems that take suction from the CST have automatic switchover to the suppression pool if CST level is low. Valves for the switchover are included in the fault trees.
The ACIWA (firewater) system (Figure 1914) is designed to injectwater into the reactor if the ECCS systems are not available. It is also the only means of water injection in case of a station blackout beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Although firewater is not a Qass IE safety system, because of the safety function described above, the firewater diesel driven pump, the firewater tank, valves, and related piping will have seismic margm above the SSE.
Because of the importance of RCIC in station blackout sequences, differences between the seismic RCIC fault tree and the internal events fault tree are explained below:
(1) The internal events fault tree contains basic events that would not be affected by an earthquake, e.g., test and maintenance unavailability. These events contribute to the random failure probability during the seismic event and are included in the random failure part of the seismic analysis. They are deleted from the RCIC seismic fault tree.
i (2) The internal events fault tree contains common-cause failure events. These l
are deleted from the RCIC seismic fault tree since a basic rule of the seismic analysis is that all like components within a system fail together.
l (3) The internal events RCIC fault tree contains separate events for the turbine and for the pump. The seismic fault tree uses a combined event, " turbine-driven pump", since that is the assembly for which there is a seismic capacity.
19I4 Seismic Margins Analysis mt
Rev.O ABWR oease conniDecamenyrien
(..
i I
I HCLPFl *RHP2, HCLPF2*RHPI, I
RHPl*RHP2, where:
the HCLPF of one event, HCLPF1
=
the random / human failure probability of that event, RHP1
=
the HCLPF of a second event, and HCLPF2
=
the random / human failure probability of the second event.
RHP2
=
The resulting combinations are reduced according to min-max rules.
191.6 Results of the Analyses The results of the convolution analysis are shown on the event trees and in Table 19I-2 in terms of HCLPF values for the accident sequences, with and without the inclusion of
)
random failures. The results of the convolution analysis in terras of accident classes are shown in Table 19I-3. The combination of HCLPF and random failure probabilities of accident sequences are described in Table 191-4.
191.7 Containment isolation and Bypass Analysis In the seismic margins analysis there were no cutsets leading to core damage with low HCLPF values. A supplemental analysis was conducted to evaluate the HCLPF values for containment isolation for events that could cause containment bypass as a result of an earthquake, with potential for large releases to the emironment.
4bowh19t-Z5 Based on the resul of the bypass analysis discussed in Subsection 19E.2.3.3 and shown on Figure 191-16 e events selected for evaluation in this analysis are:
(1) Main steam lines (Figure 19E.2-19a)
(2) Feedwater or SLC injection lines (Figure 19E.2-19b) l (3) Reactor instrument, CUW instrument, LDS instrument / sample or containment atmosphere monitoring lines (Figures 19E.2-19d,19E.2-19e, and l
19E.2-19f, respectively) i (4) RCIC steam supply or CUW suction lines (Figure 19E.2-19e)
(5) Post accident sampling lines (Figure 19E.219j)
Seismic Margins Analysis 191 8
_ _ _ _ _ _ _ _. _ _ _ _ _. _ _. _.. _ _ _ _ _ ~ _ _
I Mov. O ABWR anion cueetouanesser:
(6) Drywell sump drain line (Figure 19E.2-19j)
)
(7) SRV discharge lines (Figure 19E.2-19k)
)
(8) ECCS lines (Figure 19E.219c)
(9) Daywell inerting/ purge lines (Figure 19E.2-19i)
(10) Wetwell/drywell vacuum breaker lines (Figure 19E.2-19g)
The bypass paths for atmospheric control system crosstic lines (Figure 19E.2-19h) j require inadvertent opening of two normally closed motor operated valves. Since the seismic analysis does not consider a fail-open mode for nonnally closed valves, these l
bypass paths are not included in the analysis.
In the bypass analysis of Subsection 19E.2.3.3, several potential bypass pathways were excluded from detailed analysis on the basis of various reasons. The reasons are discussed in Subsection 19E.2.3.3.2 and Table 19E.2-1. Diese reasons were reviewed to detennine whether they remain valid in regard to seismic events. All but one of the reasons are based on configuration details that would not be affected by an earthquake.
RHR wetwell and drywell spray lines were excluded on the basis that the pipes are designed for higher internal pressures than will be seen in actual operation and would l
thus have a very low probability of breaking. In this case, the seismic event could increase the probability of a break in these lines. However, these pipes have very high seismic capaaty with very low probability of breaking due to a seismic event.
1 l
An event tree was constructed for each of the above events.These event trees are shown l
on Figures 191-16 through 191-25. All event trees start with the earthquake as the initiating event followed by a core damagmg accident. If there is no core damage there is no large release. The HCLPF and random failure probability are shown for each branch point, and the sequence HCLPFs using convolution and min-max methods are l
also shown on the figures.
Figure 191-is for suppression pool bypass via main steam lines. Following the earthquake and accident, the question is asked whether or not there is a break in a main steam line outside containment. If there is a break, the question is asked whether or not at least one MSIV in each steam line closes to isolate the break. For the case where there is no break, there could still be a bypass release to the main condenser if a turbine i
bypass valve is open-unless the MSIVs are closed to isolate the break.
i 17 j
Figure 191-MIis an event tree for bypass via feedwater or standbyliquid controllines.
These lines inject into the RPV and are protected from reverse flow by redundant check valves. These check valves provide isolation of upstream breaks provided that one of the valves closes in the line with the break.
191-9 Seismic Margins Analysis
Mev. O ABWR anew canetonementmsez
\\
l8 Figure 19I-)6is for bypass via reactor instrument, CUW instrument, LDS instrument, LDS sample or containment atmosphere monitoring lines. These lines are also protected by check valves, a single valve in each line.
19 l
Figure 19I-)6 is for bypass via either the RCIC steam supply line or the CUW suction line. Both of these lines are protected by motor operated isolation valves which require power. Since offsite power is lost due to the earthquake, emergency power is required.
Figure 191-20 is for bypass via the post accident sampling lines. These lines are also isolated by motor operated valves.
I Figure 19121 is for bypass via the drywell sump drain line. This line is protected by a motor operated isolation valve and a check valve.
i Figure 191-22 is for bypass via the SRV discharge lines. If there is a break in an SRV l
discharge line during a core-damaging accident, and that SRV is q. -n, a bypass pathway will exist. In this analysis, it is assumed that the SRV will be open during the accident.
Figure 19123 is for bypass via any of the ECCS lines. The lines of concern are the HPCF and IEFL warm-up and discharge lines. These lines are protected by motor operated isolation valves and check valves.
(.
l Figure 191-24 is for bypass via drywellinerting/ purge lives. These lines are protected by air operated valves.
Figure 19I-25 is for bypass via wetwell/drywell vacuum breaker lines. It requires an inadvertent opening of a vacuum breaker (checkvalve) to initiate a bypass during a severe accident.
I 191.8 References 191-1 R.P. Kennedy, et al., " Assessment of Seismic Margin Calculation Methods",
NUREG/CR-5270, Lawrence Livermore National Laboratory, March 1989.
I
?
I l
l I
l
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d Seismic Margins Analysis 19110 I
=
R3. 0 I
ABWR oneine ceneetoocaeeevner2 i
t.
Table 191-1 ABWR Systems and Components / Structures Fragilities (Continued) t System / Component MED_CP IAu)* LOG _STD ( c)*
HCU9 (in g)
- 12. Staney Liq. Cont. Sys. (C4)
)
- SLC Tank
- SLC Pump
-Valve (Motor Operated)
- SLC Piping
- 13. Condensate injection (V2)
Ik
- Pump (Motor Driven) l
-Injection Valve (MO)
- Piping
- Condenser Hotwell 1[. Fireweter System (FW) 3 - FW Tank'
- Pump (Diesel Driven)
- Injection Valve (Manual)
('
- FW Piping
-Valve (Manual)
t HCLPF := A,x exp(-2.326x c)
- Fireweter tank may be designed and built to a lower capacity if provision is made for a pumper truck housed in such a manner that it will survive a SSE and a hose that will reach an alternate water supply.
191-13 Seismic Margins Analysis
?
Table 190C-1 Loss of Offsite Power Precursors 3h 1
En
{
Event Description Applicable ABWR Features g
Indian Point 1 and Yankee Rowe (11/9/65)
" Great Northeast Blackout" ABWR has two independent offsite power D
sources. These are backed up by three Q
physically and electrically separate trains of a
Class 1E AC power each containing an
[
emergency diesel generator. These are further 3
backed up by a permanent onsite Combustion Turbine Generator (CTG) which is capable of y
powering any one of the three trains if all three 3
diesels were to fail. The CTG is also capable of s
supplying power to non-safety busses such that g
eedwat ondensate pumps can also be p
used to p ovide reactor coolant make up.
2 Point Beach 1 (2/501)
Loss of all transmission lines, failure of See discussion of Indian Point 2 and Yankee
{
three transformer differential relays, Rowe (11/9/65).
g causing transformer lockout.
a
?
Ginna (3/4R1)
Plant siding fellinto 34.5 kV line ABWR has two independent transformnrs connecting sole startup transformer.
powered by two independent offsite power supplies which reduces the probability of losing 1
offsite power. in the event of losing offsite i'
(
power, features described under Indian Point 2 y
and Yankee Rowe (11/9/65) can mitigate the k
event.
P11isades (9/201)
Transmission line fault, isolation breaker See discussion of Ginna (3/4/11).
failure. Backup relay isolated 345 kV bus.
a 2
San Onofre 1 (6/7#3) 138 kV auxiliary transformer out for ABWR uses three auxiliary transformers. Each g-maintenance. Ground fault operated powers one of the three Class 1E and non-1E p
differential relays, de-energizing other buses. In addition, a reserve auxiliary g
auxiliary transformers.
transformer is available to power all three Class 1E buses. CTG is also available which can P
power 1E and non-1E busses without using the j auxiliary transformers.
g a
8 3
c
=~
l SSAR CHAPTER 19 CHANGES (Except Section 19P)
I j
83A6100 Rev. 5 ABWR sundantsaferyAnalysisnevert Table 1.9-1 Summary of ABWR Standard Plant I
COL License information (Continued)
Subsection ltem No.
Subject 19.9.5 19.5 Plant Walkdowns 19.6 Confirmation of Loss of AC Power Event 19.9.6 19.7 Procedures and Training for Use of AC-independent 19.9.7 Water Addition System l
I 19.8 Action to Avoid Common-Cause Failures in the Essential 19.9.8 Multiplexing System (EMUX) and Other Common-Cause Failures 19.9.9 19.9 Action to Mitigate Station Blackout Events 19.10 Actions to Reduce Risk of Internal Flooding 19.9.10 19.11 Actions to Avoid Loss of Decay Heat Removal and 19.9.11 Minimize Shutdown Risk 19.12 Procedures for Operation of RCIC from Outside the 19.9.12 Control Room 19.13 ECCS Test and Surveillance intervals 19.9.13 19.9.14
/
19.14 Accident Management I(,,..
19.9.15 l
19.15 Manual Operation of MOVs 19.9.16 19.16 High Pressure Core Flooder Discharge Valve 19.9.17 19.17 Capability of Containment isolation Valves 19.18 Procedures to Ensure Sample Lines and Drywell Purge 19.9.18 Lines Remain Closed During Operation 19.19 Procedures for Combustion Turbine Generator @
19.9.19 pmer vency Diesel Generato)to Supply Power to Condensate Pumps 19.19a Actions to Assure Reliability of the Supporting RCW and 19.9.20 Service Water Systems 19.9.21 19.19b Housing of AICWA Equipment 19.19c Procedures to Assure SRV Operability During Station 19.9.22 Blackout 19.19d Procedures for Ensuring Integrity of Freeze Seals 19.923 l
19.19e Procedures for Controlling Combustibles During 19.9.24 Shutdown l
19.9.25 19.19f Outage Planning and Control l
19.19g Reactor Service Water Systems Definition 19.9.26 Capability huhtpakers 19.9.27 19.19h v
w COL License Information - Amendment 35 1.9-12
l E3A6100 Mov. 6
}
ABWR sandedsenyAnnwanepar l
Table of Contents (Continued) 19.6.9 Safety Goal Policy Statement..................................................................19.66 i
l.
19.6.10 Deleted........................................................................................................19.6-6 19.6.11 Co n clusio n............................................................................................... 19. 6 6 19.6.12 Re fe re n ces.................................................................................................... 19.6 19.7 PRA as a Design Tool................................................................................................. 19.7-1 19.7.1 ABWR Design and Operating Experience..................................................... 19.7-1 19.7.2 Early PRA Studi es........................................................................................... 19.7-1 i
19.7.3 PRA Studies During the Certification Effort................
........................ 19.7-3 l
19.7.4 Conduct of the PRA Evaluations................................................................... 19.7-10 l
19.7.5 Evaluation of Potential Design Improvements........................................... 19.7-10 19.8 Important Features Identified by the ABWR PRA......................................................... 19.8-1 l
19.8.1 Important Features from Ievel 1 Internal Events Analyses..........................19.8 2 19.8.2 Important Features from Seismic Analyses.................................................. 19.8 8 19.8.3 Important Features from Fire Analyses...................................................... 19.8-10 19.8.4 Important Features from Suppression Pool Bypass and Ex. Containment LOCA Analyses............................................................................................ 19. 8-13 19.8.5 Important Features from Flooding Analyses................................................ 19.816 19.8.6 Important Features from Shutdown Events Analyses...................................19.819 19.8.7 ABWR Features to Mitigate Severe Accidents.............................................. 19.8-22 19.9 CC. License I nformation............................................................................................. 19.9-1 y
19.9.1 Post Accident Recovery Procedure for Unisolated CUW Line Break...........19.91 i
19.9.2 Confirmation of CUW Operation Beyond Design Bases..............................19.9 2 19.9.3 Event Specific Procedures for Severe External Flooding............................19.9-2 19.9.4 Confirmation of Seismic Capacities Beyond the Plant Design Bases...........19.9 3 19.9.5 Plan t Walkdowns.............................................................................................. 19.9 3 19.9.6 Confirmation of Loss of AC Power Event..................................................... 19.9 4 19.9.7 Procedures and Training for Use of AC Independent Water Addition System..............................................................................................................19.94 19.9.8 Actions to Avoid Common.Cause Failures in the Essential Multiplexing System (EMUX) and Other Common-Cause Failures...................................19.9-5 19.9.9 Actions to Mitigate Station Blackout Events................................................... 19.9 5 19.9.10 Actions to Reduce Risk of Internal Flooding............................................. 19.Hi 19.9.11 Actions to Avoid Loss of Decay Heat Removal and Minimize Shutdown Risk...............................................................................................................19.9-7 19.9.12 Procedures for Operation of RCIC from Outside the Control Room........19.9-9 19.9.13 ECCS Test and Sutveillance Intervals........................................................... 19.910 l
19.9.14 Acciden t Man agemen t................................................................................ 19.9-10 l
19.9.15 Manual Operation of MOVs......................................................................... 19.9 11 19.9.16 High Pressure Core Flooder Discharge Valve............................................. 19.9-11 l
19.9.17 Capability of Containment Isolation Valves............................................. 19.9-11 l
19.9.18 Procedures to insure Sample Lines and Drywell Purge Lines Remain l
Closed During Operation...................................................
... 19.9 11 l
19.9.19 Procedures for Combustion Turbine Generator (and Emergency DieseSd Generators)to Supply Power to Condensate Pumps................................... 19.9-11 1Ni Table of Contents - Amendment 35
l 5%
m es m ev.s ABWR suudentseteryAutysis nepar Table of Contents (Continued) i 19M.8 R e fe r e n c es................................................................................
19N Analysis of Common-Cause Failure of Multiplex Equipment...................................... 19N.1 19N.1 I ntro du cti o n...........................................................................
19N.2 Re sults and Con clusions............................................................................. 19N-1 19N.3 B asis fo r th e Analysis....-.............................................................................. 19N-4 19N.4 Potential Causes of and Defenses Against EMUX CCF...............................19N.6 19N.5 Discussion cf the Effect on Core Damage Frequency..................................19N-10 19N.6 Discussion of the Effect on Isolation Capability......................................,.19N-17 l
19N.7 Summary.................................................................................................19N-18 19N.8 Re fe re n c es.......................................................................
19 0 Not Used Evaluation of Potential Modifications to the ABWR Design.......................................19P-1 19P I
.......... 19Q1 19Q ABWR Shutdown Risk Assessment............................................
19Q.1 i n tro d u c ti o n......................................................... )
19Q.2 Evaluatio n S co pe................................................................................. 19Q1 19Q.3 S u mm ary of Resul ts............................................................................ 19Q2 19Q.4 Features to Minimize Shutdown Risk...................................................... 19Q4 19Q.5 I ns tmm e ntatio n...............................................................................
19Q6 Flooding an d Fire Protection................................................................ 19Q17 19Q7 Decay Heat Removal Reliability Study..................................................... 19Q22 19Q8 U se of Freeze Seals in ABWR................................................................ 19Q 19Q.9 Shutdown Vulnerability Resulting from New Features.........................19Q35 19Q10 Procedures...........
...........................................................................19Q35 19Q.11 Summary of Review of Significant Shutdown Events: Electrical Power and Decay H eat Re moval............................................................. 19QS9 19Q.12 Results and Interface Requirements............................................... 19Q42
. 19QA-1 19QA F a u l t T re e s.........................................................................
19QB DHR Reliability Study.........
........................................................19QB-1 19QB.1 Offsite Dose and Operator Recovery Calculations........................... 19 QB 19QB.2 Time to Reach Boiling.
........................................................19QB2 19QB.3 Time for RPV Water Level to Reach Top of Active Fuel.................... 19 QB 2
.19QB-2 19QB.4 Human Reliability Analysis (HRA)..
19QB.5 Decay Heat Removal Capability of CUW and FPC.....
.19QB-4
.19QB-4 19QB.6 References..........
19 vr rable of Contents - Amendment 36 i
22AstooRev.5 ABWR standard set *y Aneinis neport 19QC Review of Significant Shutdown Evenu: Electrical Power and Decay Heat Removal.19QG1 l
19QC.1 Review of Significant Shutdown Events........................................................19QG1 l
19R Probabilisdc Flooding Analysis...................................................................................... 19R-1 1
19R.1 Introduction and Summary........................................................................... 19R-1 19R.2 Sco pe of Analysis........................................................................................... 19 R-3 19R.S Screening Analysis (Water Sources and Buildings)........................................19R-4 19R.4 Deterministic Flood Analysis.................................................,......................... 19R-4 19R.5 Probabilistic Floo d Assessm ent..................................................................... 19R-17 19R.6 Results and Interface Requirements......................................................... 19R-29 l
i
/0
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6' I
Table of Contents - Amendment 35 19-vin
~ ~ _ =. - -. -
1 23A6100 rov. 3 ABWR sanndard sdery Anotysis neport G e c) t vs 3s on a l 6 re.pk T t w -Oss (grg wAkGr 5bW COtM d re.5u mt b
86f b A h Mo o r cl v-CA n n F w o wi d !svw b l
floodsn ow Mu G Y' wder etqk4-l
\\ow Mgo/%C M Y4 Wm sf alle d e.q a s p m ew kr 4Leh bort.
thc three RCCW didsionsfrhe CDF for control building floodingis less than 1.0E-8 per P
M ad4-er d grad Joe a 4o sdeh.M*M S d M- 044<..
year.
a d n ^s 4*I ~ i y
.osu be yim Reactor building flooding could occur either inside or outside secondary containment.
In either case, the flooding sources are finite with the suppression pool and condensate
)
storage tank being the largest sources. Inside secondary containmegfl,ogiggnot cause damage to equipment in more than one of the three safety 2ivisionspecause of watertight doors on each safety division room. As was the case in the control building,
.g; 4" E= " hI ungto sump pumps on the first wder fnde.oks e er id : r.d r ;:
" I ' * ' 8
- W F oor. The available volume of rooms on the Brst floor can contain all potential flood Ec4e d' sources. Outside secondary containment, floor drains direct all flood water to sump d
dro s al pumps on floor B1F (third floor). lf the sump pumps fail or cannot keep up with the F t.. r flooding rate, an overfillline in the sumps direct water to the corddor of the first floor where it can be contained as discussed above. The CDF for reactor building flooding is approximate
-9 r year.
The total CDF for internal flooding is abo t 2.bE-8/ year for low PCHS and approximately 2.pE-8 per year for a high 9
19.4.6 ABWR Shutdown Rigk
\\
s Hodes The ABWRfesign has been evr.lua for risks associated with shutdown conditions (i.e.,(Model) 3,4 and 5). The eval tion included the following shutdown dsk h4ev clivttsona( N o ooks vwo.] U
'* 'I b b Eloor cl J %
occur (1) Decay heat removal, 6 d k k M er edevoch e Suc,b
[ 464 wo qby
- e. 4 o J o
(2) Inventory control.
eewpveavT-u di oc.cuv.
(3) Containment integnty, (
j (4) Loss of electrical power, (5) Reactivity control.
The evaluation also included risk reduction features of the ABWR due to instrumentation, flooding and fire protection, use of freeze seals, and procedure guidelines. ABWR features that are not part of current BWR designs were evaluated to determine if any new vulnerabilities would be introduced. In addition, an evaluation of approximately 200 events at operating BWR plants which were considered precursors to loss of decay heat removal capability showed that ABWR design features could mitigate the effects of all these events.
/
19A-13 Externst Event Anstysis and shutdown Misk Anstysis - Amendment 23
.- -,~ -.~.
- - ~..- - - -.-
i 23A6100 nsv. 3 A_BWR ananad u ny w nerert h.
The capability of the Reactor Water Cleanup (CUW) System to provide an additional means of decay heat removal with the reactor at high pressure wasjudged to be less important than the features selected as "important features." The additional l
redundancy prou ded by this capability does not significantly reduce the calculated ABWR CDF. Thil is due to the high reliability of other means of decay heat removal such as the various mwies of operation of the three RHRloops and the containment overpressure protection system which result in a very small contribution of Class II sequences to total CDF without the CUW capability.
i The degree of redundancy in SRVs to perforin the ADS function was alsojudged to be less important than other features. Only three SRVs are regsdred to open to depresr,urize the reactor so that low presure pumps can provide the necessary cooling.
The eight ADS SRVs plus the remaining ten SRVs that can be manually actuated far exceed redundancy requiremenu for depressurization. ADS failure is dominated by common cause failure of the ADS valves.
Another featurejudged to be less important than other features is the automatic initiation of RHR on suppression pool high temperature. Many hours are avadable to initiate RHR to remove heat from the suppression pool following transienu that dump
(
heat to the suppression pool. The reliability of operators to manually initiate this function when required isjudged to be very high, therefore this automatic initiation feature does not significantly reduce the calculated CDF.
The capability to manually inidate scram wasjudged to be less important than the selected features. The ability to manually initiate scram is not an important feature from the standpoint of CDF due to the highly reliable, redundant, and diverse features of the reactivity control systems.
The capability to use the CRD hydraulic system to provide additionalwater injection l
into the core wasjudged to be less important than the selected features. This is the l
primary reason that numerical credit for this function was not taken. The valve in some sequences for the coolant injection capability of the CRD pumps is of a lesser importance since adequate core cooling is avadable from other sources to assure a very low core damage frequency.
It was alsojudged that the high drywell pressure signal for ADS was less important than
(
the selected features. With the incorporation of the drywell high pressure signal bypass l
timer, the high drywell pressure signal for ADS is less important.
e capability to provide AC power to the condensate pumps from a safety l
was oncejudged to be an important feature. However, its importance decreased after the combustion turbine generator was added to the design. If no credit were taken for condensate pumps being powered by safety buses, core damage frequency increases by nly 6%.
~
1980 Important features identined by the ASnft MtA - Amendment 33
1 L3A6100 R;v. 5 ABWR somrardsareryAnalysisneport
\\
1 (7,
\\....
event or as a result of control blade relocation during the recovery of a badly damaged core. A possible strategy could be a caution for the operators and/or technical support staff to monitor the power level (perhaps indirectly via the rate of containment pressurization) and enter ATWS procedures as necessary.
l 19.9.15 Manual Operation of MOVs As noted in Subsection 19.7.3 (S)(a), manual operation of MOVs can be used to improve the availability of decay heat removal. The COL applicant will implement procedures for such an operation.
19.9.16 High Pressure Core Flooder Discharge Valve As noted in Subsection 19D.7.7.5, the HPCF loop B pump discharge valve is in the j
drywell. Plant procedures should include independent verification that the valve is locked-open following maintenance.
19.9.17 Capability of Containment isolation Valves To insure that containment isolation valve capability does not reduce the containment capability, the COL applicant will demonstrate that the stresses of containment isolation valves, when subjected to severe accident loadings of 0.77 MPa internal pressure and 260*C (500*F) temperature in combination with dead loads, do not exceed ASME
~
Section III service level C limits. In addition, the ultimate pressure capability at 260*C (500 F) will be shown to be at least 1.03 MPa.
19.9.18 Procedures to insure Sample Unos and Drywell Purge Lines Remain Closed During Operation As noted in Subsection 19.8.4.3, it is important that these lines be normally closed during plant operation. The COL applicant will develop procedures and adminisustive controls to ensure the valves are normally sealed closed and that the purge valves have motive power to the valve operators removed.
19.9.19 D da as for Combustion Turbine Generator (iid Emergency dim enerato@to Supply Power to Condensate Pumps The COL applicant will implement procedures for manual transfer of Combustion Turbine Generator (CTG) =d "xxgx r;"- ' O.um m.
'"") power to the condensate pumps. A rr:f b Si:ir 12E!3. ^- **M A "" : p:.:x er
- 3 ;
- :"- - -~~n :
- nz. b t: c~ '~~~~ n;:m "- 3 J., y.s1...I ::~e F'^^ *- NC r' r" r: r"N:. Condensate pump support systems (lube oil, cooling water) are also ::od Ii, y_ y. x: *^ ;xdd -
n r.. x, i ; "_" 'fr - '-- - 1 c.,2 P --
[
+k CTG.
b5ul'P P*
- 8*
19.9 11 COL License Information - Amendment 35
23A6100 Rev.1 ABWR aannsdsarnyAssiysis nspar
('
Table 19D.4-1 Bases for Core Cooling and Heat Removal Function Event Tree (Branch inputs Not Derived from Subsection 19D.6 Fault Trees)
Value Symbol Description 1.
Q Feedwater Unavailability Following a Transient 5.0E-02 (1 FW Pump + 1 Condensate Pump + 1 Cond. Transfer Pump)
It is estimated that 50% of the time feedwater pumps will trip on high water level, in the event of loss of feedwater, failure to manually recover at least one pump train is estimated to be 0.1.
2.
V2 Failure to Recover 1 Condensate and 1 Cond. Transfer Pump 1.0E-01 Similar to the preceding estimate, but at low pressure, failure to manually recover at least one pump train is estimated to be 0.1. For loss of offsite power events, a ggenerator bus transfer is required and assumed.
6Com 5
0" 3.
W1 Normal Heat Removal (NHR) 1.0E-02 The NHR failure probability of 1.0E-02 is taken from Section D.1.5 of GESSAR and is a conservative application due to the improved reliability expected from the use of motor-driven feedwater pumps.
1.0E-01 4.
W2 Reactor Water Cleanup System (CUW)
The CUW System is capable of removing decay heat at high RPV pressures if retum water bypasses the regenerative heat exchanger. Failure to manually activate this alternate heat removal system is taken to be 0.1.
Table 19D.4-2 Event Tree Branch Point Values (Not Derived From Subsection 190.6 Fault Tree)
Symbol Description Referenes Value Tm Event Frequency Table 19D.3-1 1.0579 (1.0+
Transfer from Fig.190.4-4)
Q Failure to inject with 'eedwater Table 190.4-1 5.0E-02 U2 Failure to inject with ccndensate Table 190.4-1 1.0E-01 W1 Failure to restore normal heat removal Table 190.4-1 1.0E-02 W2 Failure to actuate CUW Table 190.4-1 1.0E-01 190.4-7 Accident Event Treen - Amendment 31
mean. a ABWR seadedsanyaanysiaerert I
l l
I high-pressure injection systems fail, the reactor must be denra="+-d ranidly for low-i pressure system use (LPFL Vl)}r :::d:nn; Ljcam.. -Vb L A",V'", cEr.dM j
p--~ ~ be r d: :: r % --- g-g b~ by *L an-m. T..: mq I
a ie n is o g j
191.3.3 ATWS Event Tree Figure 191-3 (ATWS transfer) represents failure to scram, and requires standby liquid control (automatic) and operator action to control reactor water level with the injection l
system (s) that are available. The HEP for this action is 0.01. In this ATWS analysis, if high-pressure systems fail, core damage resulu. No credit is given to low-pressure l
injection. For an ATWS, the probability of a stuck + pen SRV was conservatively l
increased to 0.1, on the basis ofincreased SRV activity.
191.4 System Analysis The fault trees used in the seismic system analysis are shown on Figures 191-4 through j
l 191-15. The seismic system analysis calculates the probability of seismic failure and I
corresponding system HCLPFs of each of the important systems throughout the seismic ground acceleration spectrum. The system HCLPFs are then input to the event trees f
and combined with random system failure probabilities and human errors.The seismic l
fault trees contain only those components that might be subject to seismic failure.
l Random system failure probabilities are taken from the internal events analysis and l
include all other emnponents. One of the important ground rules of the seismic l
margins analysis is that all like components in a system always fail together.
The reactor protection system, control rod drive system, and alternate rod insertion system were not modeled since the failure of control rods to insert is dominated by the relatively low seismic fragility of the fuel assemblies, control rod guide tubes, and housings. A seismic fault tree for reactivity control is shown on Figure 191-13. The fuel l
assemblies are the most fragile component.
A seismic fault tree for the standby liquid control system is shown on Figure 191-14.
Failure of the standby liquid control system is dominated by failure of two components:
the pump and boron supply tank.
l Since the most fragile essential component in the plant is the ceramic insulator in the switchyard, the loss of offsite power dominates the analysis and the avadability of emergency power becomes very important. The loss of-power fault tree (Figure 191-10) is for emergency AC power. In the loss of emergency AC power fault tree, the more fragile components are the diesel generator, transformers, motor control centers, inverter and circuit breaker. The DC power fault tree (Figure 191-11) has two elements:
batteries and cable tray.
1944 Seismic Margins Analysis - Amendment 34 i
I 23A6100 R3v. 4 ABWR suasnt setery Analysisnepart l
cw A c.ow h%dts_ Nt tA M udem Systems and equipment which require offsite power, such as the feedwater systenikare not modeled since offsite power is presumed to be not available for the core damage sequences.pe condensate injection system is modeled on Figure 191-15 since credit is given to the operator for transferring condensate to an emergency bus (Figure 191-2.)/
4 The human error probability (0.1) is much greater than the seismically induced equipment failure probability, therefore, this fault tree has negligible impact on the J HCLPF value of the corresponding accident sequencesj Essential service water is as important as emergency power, and its loss would have much the same effect as the loss of emergency AC power. The lossef-service-water fault tree is shown on Figure 191-12. The more fragile components in this system are the service water pump, heat exchanger, and room air conditioning unit. The service water pump house, with a HCLPF of 0.60, is also included in this fault tree.
Structure failures that could contribute to seismic core damage are shown on Figure 19I-9. In this analysis, any one or more of these structural failures are l
conservatively presumed to result in core damage. The structures having the lowest sei smic capacity are the reactor building and control building.
The remainder of the fault trees are for core cooling (Figures 19I-4 through 191-8). The more fragile components in these systems are the pumps, heat exchangers, and the firewater supply tank. The condensate storage tank (CST) is not modeled since the ECCS systems that take suction from the CST have automatic switchover to the suppression pool if CST level is low. Valves for the switchover are included in the fault trees.
L The ACIWA (firewater) system (Figure 191-8) is designed to inject water into the reactor if the ECCS systems are not available. It is also the only means of water injection in case of a station blackout beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Although firewater is not a Class 1E safety system, I
because of the safety function described above, the firewater diesel driven pump, the l
firewater tank, valves, and related piping will be seismically designed for a HCLPF of l
l 0.5 g.
Because of the importance of RCIC in station blackout sequences, differences between the seismic RCIC fault tree and the internal events fault tree are explained below:
l (1) The internal events fault tree contains basic events that would not be affected by an earthquake, e.g., test and maintenance unavailability. These events contribute to the random failure probability during the seismic event and are included in the random failure part of the seismic analysis. They are deleted from the RCIC seismic fault tree.
191-5 Seismic Margins Analysis - Amendment 34
i l
23A6100 R:v. 4 ABWR samwedsarnyAnalysisnoper l
HCLPF values for containment isolation for events that could cause containment bypass as a result of an earthquake, with potential for large releases to the environment.
l Based on the results of the bypass analysis discussed in Subsection 19E.2.3.3 and shown on Figure 19116 the events selected for evaluation in this analysis are:
ro*p 191-fA (1) Main steam lines (Figure 19E.2-19a)
(2) Feedwater or SLC injection lines (Figure 19E.2-19b) 1 (3) Reactor instrument, CUW instrument, LDS instrument / sample or containment atmosphere monitonng lines (Figures 19E.2-19rl,19E.2-19e, and 19E.2-19f, respectively)
(4) RCIC steam supply or CUW suction lines (Figure 19E.2-19e)
(5) Post accident sampling lines (Figure 19E.2-19j)
(6) Drywell sump drain line (Figure 19E.2-19j)
(7) SRV discharge lines (Figure 19E.2-19k)
(8) ECCS lines (Figure 19E.2-19c)
(9) Drywell inerting/ purge lines (Figure 19E.2-19i)
(10) Wetwell/drywell vacuum breaker lines (Figure 19E.2-19g)
The bypass paths for atmospheric control system crosstie lines (Figure 19E.2-19h) require inadvertent opening of two normally closed motor operated valves. Since the seismic analysis does not consider a fail-open mode for normally closed valves, these bypau paths are not included in the analysis.
In the bypass analysis of Subsection 19E.2.3.3, several potential bypass pathways were l
excluded from detailed analysis on the basis of vadous reasons. The reasons are i
discussed in Subsection 19E.2.3.3.2 and Table 19E.2-1. These reasons were reviewed to I
determine whether they remain valid in regard to seismic events. All but one of the reasons are based on configuration details that would not be affected by an earthquake.
RHR wetwell and drywell spray lines were excluded on the basis that the pipes are l
designed for higher internal pressures than will be seen in actual operation and would thus have a very low probability of breaking. In this case, the seismic event could increase the probability of a break in these lines. However, these pipes have very high l
seismic capacity (3.0 g) with very low probability of breaking due to a seismic event.
An event tree was constructed for each of the above events. These event trees are shown on Figures 191-16 through 191-25. All event trees start with the earthquake as the Seismic Margins Analysis - Amendment 34 1 91-9
23A6100 RJv. 4 ABWR sendantsarnyAnairsisnopear L
initiating event followed by a core-damaging accident. If there is no core damage there is no large release. The HCLPF and random failure probability are shown for each branch point, and the sequence HCLPFs using convolution and min-max methods are also shown on the figures.
Figure 191 s for suppression pool bypass via main steam lines. Following the earthquake and accident, the question is asked whether or not there is a breakin a main steam line outside containment. If there is a break, the question is asked whether or not at least one MSIV in each steam line closes to isolate the break. For the case where there is no break, there could still be a bypass release to the main condenser if a turbine bypass valve is open-unless the MSIVs are closed to isolate the break. The two bypass l
sequences for this event both have min-max HCLPF capacities of 0.74 g.
t 17 Figure 191Jefis an event tree for bypass via feedwater or standby liquid controllines.
These lines inject into the RPV and are protected from reverse flow by redundant check valves. These check valves provide isolation of upstream breaks provided that one of the valves closes in the line with the break. The two bypass sequences for this event also have l
min-max HCLPF capacities of 0.74 g.
LS Figure 19I-)4fis for bypass via reactor instrument, CUW instrument, LDS instrument, LDS sample or containment atmosphere monitoring lines. nese lines are also protected by check valves, a single valve in each line. The bypass sequence for this case i
l also has a min-max HCLPF of 0.74 g.
)
19 Figure 191J4fis for bypass via either the RCIC steam supply line or the CUW suction line. Both of these lines are protected by motor operated isolation valves whien require power. Since offsite power is lost due to the earthquake, emergency power is required.
l The two bypass sequences for this event both have min-max HCLPFs of 0.74 g.
Figure 191-20 is for bypass via the post accident sampling lines. These lines are also isolated by motor operated valves. The bypass sequences for this event also have min-l max HCLPFs of 0.74 g.
Figure 191-21 is for bypass via the drywell sump drain line. This line is protected by a motor operated isolation valve and a check valve. Both components have HCLPF l
capacities of 0.74 g and the two bypass sequences have min-max HCLPFs of 0.74 g.
Figure 191-22 is for bypass via the SRV discharge lines. If there is a break in an SRV discharge line during a core-damaging accident, and that SRV is open, a bypass pathway will exist. In this analysis, it is assumed that the SRV will be open during the accident.
The resulting HCLPF capacity for this sequence is the capacity of the SRV discharge line l
(0.74 g).
1N-10 Seismic Morpins Analysis - Amendment 34
~.
I 23A6100 n:v. 3 ABWR sentertsareryAnalysis nopers Table 191-1 ABWR Systems and Components / Structures Fragilities (Continued)
System / Component MED_CP (Au)
LOG _STD ( c)
HCLPF* (in g)
- 12. Standby Liq. Cont. Sys. (C4) i
- SLC Tank 1.8
.46
.62
- SLC Pump 1.8
.46
.62
- Valve (Motor Operated) 3.0
.60
.74 j
l
- SLC Piping 3.0
.60
.74
/
TCondensato injection (V2)
{
- Pump (Motor Driven) 1.8
.46
.62
- Injection Valve (MO) 3.0
.60
.74
- Piping 3.0
.60
.74 Condenser Hotwell 2.1
.45
.74 t
G g Firewster System (FW) l
- FW Tank 1.5'
.46
.51 l
- Pump (Diesel Driven) 1.5
.46
.51 l
-Injection Valve (Manual) 2.0
.60
.50 l
FW Piping 2.0
.60
.50
[
- Valve (Manual) 2.0
.60
.50
- HCLPF = A, x exp (-2.326xS,)
t Firewster tank may be designed and built to a lower capacity if provision is made for a l
pumper truck housed in such a manner that it will survive a SSE and a hose that will reach an alternate water supply.
I
{
l t
19114 Seismic Margins Analysis - Amendment 33
l 23A6100 C2v. 3 ABWR senderdsareryAur eisneuen r
Table 1912 Seismic Margins for ABWR Accident Sequences (Convolution Method)
With Random Failure Without Random Failure Accident Sequence HCLPF MED_ CAP LOG _STD HCLPF MED CAP LOG _STD Number *
(ing)
(A )
(S,i (in g)
(A )
($,)
m l
1 0.64 1.14 0.25 0.64 1.14 0.25 l
2 0.89 2.02 0.35 0.89 2.02 0.35 l
3 0.81 3.00 0.56 0.81 3.00 0.56 4
1.21 3.34 0.44 1.21 3.34 0.44 l
5 0.77 1.40 0.26 0.82 1.43 0.24 l
6 1.02 2.09 0.31 1.03 2.10 0.30 l
7 0.98 3.01 0.48 0.99 3.01 0.48 8
1.29 3.34 0.41 1.29 3.34 0.41 l
9 0.73 1.23 0.23 0.73 1.23 0.23 l
10 0.94 2.01 0.33 0.94 2.01 0.33 l
11 0.77 2.37 0.48 0.77 2.37 0.48 12 1.21 2.79 0.36 1.21 2.79 0.36 13 1.02 2.30 0.35 1.02 2.30 0.35 14 1.33 2.65 0.30 1.33 2.65 0.30 e,o39 1,gO O 2.6 1.0 %
15 W
W 6
% t.1) W o. 2,_c;
+94--
16 1.13 3.04 0.43 1.14 3.04 0.42 17 A
g.-3EF-y y
p 18 1.46 4.16 0. u 0.45 6
.5 E'D 1.46 4.16 D O.45 0
19 0.96 1.68 0.24 0.97 1.69 0.24 20 0.89 3.00 0.52 0.90 3.00 0.52 21 0.95 2.82 0.47 1.08 3.04 0.44 22 1.26 4.05 0.50 1.39 4.16 0.47 23 0.87 2.98 0.53 0.90 3.00 0.52 24 0.80 1.44 0.25 0.80 1.44 0.25 25 0.96 1.09 0.24 0.97 1.69 0.24
- See event trees.
Seismic Mergins Analysis - Amendment 33 19115
23A6100 Cev. 3 ACWR standard setery Analysis neport Table 191-3 Seismic Margins for ABWR Accident Classes (Convolution Method)
With Random Failure Without Random Failure Accident HCLPF MED_ CAP LOG _STD HCLPF MED_ CAP LOG _STD Class (in g)
(A )
(e (in g)
(A l (Sel m
l lA 0.75 1.68 0.35 0.76 1.68 0.34 l
18 2 0.64 1.14 0.25 0.64 1.14 0.25 IC 0.90 1.44 0.20 0.92 1.46 0.20 g,g g,
l ID 0.78 0.82 W
l IE 1.02 2.30 0.35 1.02 2.30 0.35 l
IV 0.70 1.13
. 0.21 0.71 1.14 0.20 l
lA P, IE P 0.89 1.46 0.22 0.89 1.47 0.21 i
i l
i Seismic Margins Analysis - Amendment 33 19I.16 i
swamv. s ABWR aondedsdoryAnnopeinnever Table 191-4 HCLPF Derivation for the ABWR Accident Sequences (MIN-MAX Method) l Sequence 1
- APW'FA -+ (0.60g+1.6E-3)*(0.50g+1.0E-3) -*
l
- 0.60g, 0.50g
- 1.6E-3 l
Sequence 2
- 0.70g*(0.60g+1.6E 3)'(0.50g+1.0E-3) -+
- 0.70g Sequence 3
- X*APW -+ 0.74g*(0.60g+1.6E 3) -+
- 0.74g Sequence 4
- HX'X'APW -+ 0.70g*0.74g*(0.60g+1.6E 3) -+
- 0.74g l
Sequence 5
- FA*UR'APW -+ (0.50g+1.0E-1)*(0.70g+6.0E-2)*(0.60g+1.6E-3) -+
l
- 0.70g,0.60g*6.0E 2 Sequence 6
- HX*FA*UR*APW -+
l
- 0.70g*(0.50g+1.0E-1)*(0.70g+6.0E-2)*(0.60g+1.6E 3) -+
- 0.70g Sequence 7
- X'UR*APW -+ 0.740g'(0.70g+6.0E 2)*(0.60g+1.6E 3) -+
- 0.74g Sequence 8
- HX*X'UR'APW -+ 0.70g*0.74g*(0.70g+6.0E-2)*(0.60g+1.6E 3) -+
- 0.74g Sequence 9
- C*APW -* 0.62g*(0.60g+1.6E-3) -+
- 0.62g Sequence 10
- HX'C'APW -+ 0.70g*0.62g'(0.60g+1.6E 3) -+
- 0.70g Sequence 11
- DP -+ 0.74g
- 0.74g Sequence 12
- HX'DP -+ 0.70g*0.74g
- 0.74g Sequence 13
- Si *
- 1.11g Sequence 14
- 1.11g Sequence 15 1'UH'UR -+
Q62g+1.0E-1).62g *(0.62g+2.7E-3)*(0.70g+6.0E-2) -*
- 0.70s,0.62g*6.0E 2 seismic Morains Analysis - Amendment as 101 11
23A6100 R:v. 2 ACWR standardsarery Analysis neport Table 191-4 HCLPF Derivation for the ABWR Accident Sequences (MIN-MAX Method) l Sequence 16
- X'UH'UR -+ 0.74g*(0.62g+2.7E-3)'(0.70g+6.0E 2) -+
- 0.74g l
Sequence 17
- 2 1'UH'PC -+
- Q.62g+1.0E-1).62 g * (0.62g + 2.7 E-3)* (0.74g +2.0E-3) -+
l
- 0.74g,0.62g*2 OE 3 l
Sequence 18
- X'UH'PC -+ 0.74g*(0.62g+2.7E-3)*(0.74g+2.0E-3) -+
l
- 0.74g 1
l Sequence 19
- UR'UH'C -+ (0.70g+6.0E-2)*(0.629<2.7E-3)*0.62g -+
l
- 0.70g,0.62g'6.0E 2 l
Sequence 20
- PA*C * (0.74g+2.4E 3)*0.62g -+
i
- 0.74g, 0.62g*2.4E-3 l
Sequence 21
- UH'PC1*C -+ (0.62g+2.7E-3)*(0.74g+1.0E 1)*0.62g -+
1
. 0.74g,0.62g*1.0E 1 l
Sequence 22
- PA*PC1'C -+ (0.74g+2.4E-3)*(0.74g+1.0E 1)'O.62g,
- 0.74g l
Sequence 23
- LPL*C -+ (0.74g+1.0E 2)*0.62g -+
l
- 0.74g, 0.62g*1.0E-2 l
Sequence 24
- C4*C -+ (0.62g+1.4E-2)*0.62g -+
- 0.62g l
Sequence 25
- UH'C4*C -+ (0.62g+2.7E 3)'(0.62g+1.4E-2)*0.62g -+
l 0.62g l
l t
i 191-18 Seismic Margins Anstysis - Amendment 32 l
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23A6100 R:v.1 Standard SafetYAnalysis Report 10 A
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Seismic Marpins Analysis - Amendment 31 191-33
I
^
i i
l l
3h
?
Tatdo 190C-1 Loss of Offsite Power Precursors W
( Event Description Applicable ABWR Features g
O l
g
$ indian Point 2 and Yankee Rowe (11/9/65)
" Great Northeast Blackout" ABWR has two independent offsite power sources.These are backed up by.three physically and electrically separate trains of 9;
g Class 1E AC power each containing an I
a emergency die.;el generator.These are further
[
backed up by a permanent onsite Combustion Turbine Generator (CTG) which is capable of 5
powering any one of the three trains if all three y
diesels were to fall. The CTG is also capable of 6
supplying power to non-safety busses such that s
Weedwater opondensate pumps can also be R
used to provide reactor coolant make up.
g Loss of all transmission lines, failure of See discussion of Indian Point 2 and Yankee Point Beach 1 (2/5/71) u three transformer differential relays, Rowe (11/9/65).
f causing transformer lockout.
g Plant siding fellinto 34.5 kV line ABWR has two independent transformers connecting sole startup transformer.
powered by two independent offsite power i
l[
Ginna (3/4/71)
,{
supplies which reduces the probability of losing offsite power. in the event of losing offsite power, features described under Indian Point 2 g
and Yankee Rowe (11/9/66) can mitigate the event.
I Transmission line fault, Isolation breaker See discussion of Ginna (3/4/71).
g Palisades (9/2/71) failure. Backup relay isolated 345 kV bus.
l;a; 138 kV auxiliary transformer out for ABWR uses three auxiliary transformers. Each lR San Onofre 1 (6/7/73)
A maintenance. Ground fault operated powers one of the three Class 1E and non-1E I
differential relays, de-energizing other buses. In addition, a reserve auxiliary transformer is available to power all three Class auxiliary transformers.
1E buses. CTG is also available which can power IE and non-1E busses without using the auxiliary transformers, g.
4 b
!L 2
ATTACHMENT 3 DCD/SSAR Pages Affected by the Reference 3 Proposed Changes (in addition to those in Reference 4)
Note: All DCD (Tier 1 and Tier 2) changes are included. The CDM and SSAR l
changes are the same except as included.
i
The pages in this Attachment 3 are assembled as they are related to change packages included in Reference 3 [ Letter (MFN 050-96) fromJoseph F. Quirk (GE) to Dennis M. Crutchfield (NRC), dated April 16,1996].
Change Package No.
Title CP1 Eliminate Hot Water Heating for RB and RWB HVAC and Reconfigure RB HVAC Filters CP 2*
Additional Chiller / Pump Set CP3 Change of Smoke Removal Method and Duct Connections CP 4*
Reassign MCR Exhaust Fan Designations CP5 Miscellaneous Tier I and Tier 2 Changes CP6 Independence of Power for Each Pair of Motor-Operated Isolation Dampers CP 7**
Eliminate RCIC Exhaust Bypass Line and Rupture Disks CP 8**
Increase FMCRD Scram Piping Design Pressure CP 9**
Use Higher Strength Material for Cladded Shells of RPV Pedestal and Tunnels CP 10 Changes to Technical Specifications CP 2 is combined with CP 4 in this attachment because markups involve same drawings.
- There are no pages associated with these three change packages.
i Rev.o
_ABWR oesten CondDocuantmer1
(
2.11.3 Reactor Building Cooling Water System Design Description l
The Reactor Building Cooling Water (RCW) System distributes cooling water through j
three physically separated and electrically independent divisions. The system removes l
heat from plant auxiliaries and transfen it to the Ultimate Heat Sink (UHS) via the l
Reactor Service Water (RSW) System. The RCW System removes heat from emergency l
core cooling equipment, induding the emergency diesel generators (DGs) during a j
safe reactor shutdown cooling function. RCW System configurations are shown in l
Figures 2.11.Sa, 2.11.3b, and 2.11.Sc. Figure 2.11.3d shows the RCW System control l
interfaces. All components cooled by the RCW System are parts of other systems and are I
not part of the RCW System. Each RCW divtsion includes two pumps which circulate i
cooling water through the equipment cooled by the RCW System and through three heat exchangers which transfer the RCW heat to the UHS via the RSW System.
The RCW System performs a safe reactor shutdown cooling function following either a l
loss-of-coolant accident (LOCA) or a loss of-preferred-power (LOPP)or both. Assuming l
a single active failure in any mechanical or electrical division or RCW support system, j
which disables any one of the three RCW divisions, the other two divisions perform safe l
(
reactor shutdown cooling.
Tables 2.11.3a,2.11.3b, and 2.11.Sc show which equipment receives RCW flow during I
various plant operating and emergency conditions. The tables also indicate how many i
heat exchangers are in service under each condition.
The RCW System is classified as safety-related except for those portions as shown on Figures 2.11.3a, 2.11.3b, and 2.11.Sc as non-nuclear safety.
The RCW System responses to a LOCA signal are the following-(1) Stans any standby RCW pumps.
(2) Opens any closed standby RCW heat exchanger outlet valves.
(3) Opens all Residual Heat Removal (RHR) System heat exchanger cooling l
water outlet valves.
(4) Closes all RCW containment isolation valves.
i l
(5) Closes valves to the following non-safety-related components (to Reactor Water Cleanup System (CUW) =d Mn ?! a.- ::--t;; (*H) S;-~a % <
% and reactor internal pump (RIP) MG sets).
'Ich Reactor Building Cooling Water System 2.11 3*1
My. 0 ABWR outen coneetouementmer2
(
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Safety Closel-Require-Seismic d
m ent*
Category' Notes Principal Component
- Class" Lacetion' fication E
6.
Other non-safety-N SC,RZ,X 1
related electrical components E
P14 Heating Steam and N
T,SC,W Condensate Water Return Syste m 6tet "T "
l E
P15 House Sollee N
T l
CPI P16 Hot Water Hosting System N
X E
6ple.
c-l i
P17 Hydrogen Water Chemistry N
T E
System l
P18 Zincinjection System N
T E
l l
E P19 Sreathing Air System N
C,SC,T E
l P20 Sempling System (includes N
SC,RZ,T l
PASS)
E P21 Freeze Protection System N
O E
P22 fron injection System N
T R1 Electrical Power Distrbution System B
l 1.
120 VAC safety-related 3
SC,X, distribution equipment RZ,U including inverters B
l 2.
Safety related Motors 3
SC,C,X, RZ,U Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.24 OsssiRcation of Structures, Components, and Systems
m..
i Rev.0 l
_ouien contretoecamentmer2
!\\
Table 6.2-9 Secondary Containment Penetration List" (Continued)
Penetration Name Elevation Diameter Number (mm)
(mm) 32 RCW (C)
-1700 100 33 RCW (C)
-1700 200 34 RCW (C)
-1700 200 35 HS 4800 150 36 MS 4800 80 37 LCW (FPC) 4800 150 i
38 LCW (CUW) 4800 150 39 RCIC 4800 50 l
40 MS (4) 16191 700 j
41 FDW (2) 13810 600 t
42 HVAC Exhaust 27200 t
43 HVAC Supply 31700 44 Controlled Access (2) 12300
(
45 Equipment Lock 12300 "5"#
I 8
46 Railroad Car Door 12300 47 HS 12300 150 48
"?!"
1x-:~;
1:;0
-r cgg i
49 1;.~;,
100 50 HNCW 12300 200 l
51 HNCW 12300 200 95 52 MUWP 4800 150 l
% e3ed 55
-l"'"l 5G
( d Del C +"
56 l l" " l O^^0 50 o'
57 Cabletrays 23500 f
58 Cabletrays 12300 59 Cabletrays 4800
- This tatie is provided in response to Question 430.34, t These HVAC openings have safety-related isolation valves with both local monitoring and remote (in control room) monitoring.
- These doors are monitored in the control room as per Subsection 13.6.3.4.
6.2-183 Containment Systems
_. --_ ~- -
l i
Riv. 0 ABWR Desen connt0*cannawa2 j
(
Table of Contents (Continued) 9A.4.1.6.2 2 Not Used................................................................ 9A.4-29 7 9A.4.1.6.23 No t Used............................................................... 9A.4-29 7 9A.4.1.6.24 Upper D/G A HVAC Room (Rm No. 653)............... 9A.4-297 9A.4.1.6.25 FMCRD A/C Panel Room (Rm No. 654)..................... 9A.4-298.
9A.4.1. 6.26 Not Used.................................................................... 9A.4-300 1
9A.4.1.0.27 Not Used....................................................................... 9A.4-300 9A.4.1.6.2 8 Not Used.................................................................. 9A.4-300 9A.4.1.6.29 Not Used.............................................................-.... 9A.4 300
)
9A.4.1.6.30 Upper D/G C HVAC Room (Rm No. 673)................. 9A.4-300
-I 1
9 A.4.1.6.31 No t Used................................................................ 9A.4-302 9A.4.1.6.32 Upper D/G B HVAC Room (Rm No. 663)................. 9A.4-302 9A.4.1.6.33 Upper Corridor B (Rm No. 626)............................... 9A.4-304 9A.4.1.6.3 4 No t Used............................................................. 9A.4 306 9A.4.1.6.35 FMCRD D/B Panel Room (Rm No. 681)..........-.......... 9A.4-306 -
9A.4.1.6.36 Not Used............................................................. 9A.4 307 9A.4.1.6.3 7 Not Used................................................................. 9A.4 307 I
9A.4.1.6.38 MS Tunnel HVH Room (Rm No. 685)....................... 9A.4 307 9A.4.1.6.39 Pits and Pools.............................................................. 9A.4-309 9A.4.1.6.40 PVC Purge Exhaust Fan (Rm No. 623)..................... 9A.4-310 9A.4.1.6.41 D/G C Corridor Room (Rm No. 635)........................ 9A.4-311 l
_ 9A.4.1.6.42 _ RIP Power Supply Room (Rm No. 638)...................... 9A.4-312 l
g_
( Rm No. 640 )....................................... W......"....
'..........g w.s. r 9A.4.1.6.43 a mm..
..._,..m n..__.--- = -- eer Room e
-u 9A.4-314 C P1.1 y.
9A.4.1.6.44 Fission Product Monitoring (Rm No. 657)............... 9A.4-315 c, l
9A.4.1.6.45 Room No. 658.......................................................... 9A.4-317 -
1 l
9A.4.1.6.46 Containment Atmospheric Monitoring System
.........b......... 9A.4-318 (CAMS) Rack A (Rm No. 659).............
' d "M
,.o 9A.4.1.6.47 w"...."... "... ""-. _,"..".. ". -. Room (Rm No. 680).............................................................. 9A.4-320 9A.4.1.6.48 Not Used................................................................ 9A.4-3 22 9A.4.1.6.49 Containment Atmospheric Monitoring System (CAMS) Rack B (Rm No. 621)................................ 9A.+322 j
9A.4.1.6.50 N ot Used............................................................... 9A.4-3 24 9A.4.1.7 Building - Reactor Bldg EI 31700mm............................. 9A.4-324 l
9A.4.1.7.1 Reactor Building Operating Deck (Rm No. 716).......... 9A.4 324 j
i 9A.4.1.7.2 RIP (A) Supply Fan and RCW (C) Surge Tank (Rm No. 715)......................................................... 9A.4-326 9A.4.1.7.3 No t Used.........................:......................................... 9A.4-328 9A.4.1.7.4 DG (C) Exhaust Fan Room (Rm No. 730)............... 9A.4-328 i
9A.4.1.7.5 Not Used.................................................................... 9A.+329 9A.4.1.7.6 RIP (B) Supply Fan and RCW (B) Surge Tank
( Rm No. 740).............................................................. 9A.4-329 9A.4.1.7.7 Access Service Area (Rm No. 764)................................. 9A.4-331
]
9A.4.1.7.8 Refueling Machine Control Room (Rm No. 760)......... 9A.4-333 9A.4.1.7.9 Gallery (Rm No. 762)................................................. 9A.4 334 9A.4.1.7.10 Memnhe Cortidor (Rm No. 761).......................... 9A.4-336 9A.4.1.7.11 Roof A/C Area (Rm No. 810 and 830).................... 9A.4-337 Table ofContents SA.H l
t
l l
l R:v.0 ABWR onien controscocwnontmer2 (S) Radioactive Material Present-None that can be released as a result of fire.
(4) Qualifications of Fire Barrien-The exterior wall,inside wall, ceiling and floor of this corridor are of S h fire-resistive construction. This corridor extends across the reactor building. At the south end of the corridor, a 3 h fire-l resistive door opens to the I".y h= =d=gn rd p;=p r:f(Rm 640). c P d.
l Atthe other end of the corridor, a nonrated door opens into D/G (A) exhaust fan area (Rm 613),
g,,.g a room
(,
9 g, (5) Combusdbles Present:
l Fire Loading Total Heat of Combusdon (MJ) 2 2
None 727 MJ/m NCLL (727 MJ/m maximum average) applies.
(6) Detection Provided -Class A supervised POC in the room and manual alarm j
pull station at Col.1.0-B.2 and 6.2-B.O.
(7) Suppression Available:
Type Locadon/Actuadon Standpipe and hose reel Col.1.0-B.2 & 6.2-B.0/ Manual ABC hand extinguishen Col.1.0-B.2 & 6.2-B.0/ Manual l
(8) Fire Protection Design Criteria Employed:
(a) The function is located in a separate fire resistive enclosure.
(b) Fire detection and suppression capability is provided and accessible.
(c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers.
(9) Consequences of Fire--Alternate routes to the areas interconnected by the corridor are provided.
Smoke from a fire will be removed by the EHVAC(A) system operating in its smoke removal mode.
(10) Consequences of Fire Suppression--Suppression extinguishes the fire. Refer to Section S.4 " Water Level (Hood) Design", for the drain system.
Analysis 9A.4-264
Rev. O ABWR onka cammcanannier2
(
(2) Equipment See Table 9A.6-2 i
Safety-Related Provides Cm Cooling Yes, D1 Yes, D1 (3) Radioactive Material Present-None that can be released as a result of fire.
(4) Qualifications of Fire Barriers-All walls are of S h fire-resistive construction.
A section of the ceiling below the FMCRD panel room (Rm 681, fire, F7200) is of S h fire-resistive concrete construction. Sections of the floor above the D/G Fan and HVAC room B (Rm 524), and the service corridor B (Rm 527) are of S h fire resistive concrete construction. Access to the area is provided from the stairs and elevator (areas $29 and 328 respectively), from corridor cpi Rm 635 (via corridor Rm 626) and from the("?" h= c el r.g;. nr.
M/44 (Rm 640). Each access route is through a S fire-resistive door.
ette M cal
- 1." *D ^ " * *
(5) Combustibles Present Fire Lc+g Totti Heat of Combustion (MJ) 2 2
Cable Tray 727 MJ/m NCLL (727 MJ/m maximum average) applies
)
(6) Detection Provided-Class A supervised POC in the room and manual alarm pull stations at 1.4 D.7,1.0-B.2.
(7) Suppression Available:
Type Locadon/Actuadon Standpipe and hose reel Col.1.4-D.7, and 1.0-B.2/ Manual ABC hand extinguishers Col.1.4-D.7,and 1.0-B.2/ Manual (8) Fire Protection Design Criteria Employed:
(a) The function is located in a separate fire-resistive enclosure.
(b) Fire detection and suppression capability is provided and accessible.
(c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers.
l 9A.4-287 Analysis
an. o oess a cornrotoocanoenvrier2 ABWR u
(.
elech cal ep;pms (3) Radioactive Material Present-None.
(4) Qualifications of Fire Barrien-The walls common with the "?S pu.T.p ud pg g
her 9 ;;: room (Rm 640), the SBGTS filter train room (Rm 642),
corridor room (Rm 614), the floor above the steam tunnel and the ceiling
(,f4/94 serve as fire baniers between adjacent fire areas and are of 3 h fire-resistive concrete constmction. A 3 h fire rated door provides access from the AC filter / fan area (Rm 615). Room 643 connects directly into room 622.
(5) Combustibles Present:
\\
Mre Loading Total Heat of Combusdon (MJ) 2 2
Cable Tray 727 MJ/m NCLL(727 MJ/m maximum average) applies j
(6) Detection Provided-Class A supervised POC in the room and manual alarm pull stations at 2.7-C.0 and 2.8-F.1.
(7) Suppression Available:
Type Locadon/Actuadon Standpipe and hose reel Col. 2.7 C.0,& 2.8-F.1/ Manual ABC hand extinguishers Col. 2.7-C.0,k 2.8-F.1/ Manual l
(8) Fire Protection Design Criteria Employed:
(a) The function is located in a separate fire-resistive enclosure.
(b) Fire detection and suppression capability is provided and accessible.
(c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers.
(9) Consequences of Fire-The postulated Sre assumes the loss of the function.
Loss of the SGTS by an exposure fire is acceptable.
Smoke from a fire will be removed by the normal HVAC System operating in its smoke removal mode.
(10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4 " Water Level (Flood) Design", for the drain system.
Analysis 9A.4-202
R:v. o ABWR oesien cenereroecamenttrier2 I
(11) Design Criteria Used fc,r Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System:
(a) Provision of raised supports for the equipment (b) Refer to Section 3.4 " Water Level (Flood) Design", for the drain system.
(c) ANSI BSI.1 standpipe (rupture unlikely)
(12) Fire Containment or Inhibiting Methods Employed:
(a) The functions are located in a separate fire resistive enclosure.
(b) The means of fire detection, suppression and alarming are provided and accessible.
(13) Remarks-None.
SA.4.1.6.20 SGTS B Division 2 Room (Rm No. 641)
(1) Fire Area-F4201 (2) Equipment: See Table 9A.6 2 Safety-Related Provides Core Cooling Yes, D2 No (3) Radioactive Material Present-Filters within their housing may become contaminated with use. Releases up the stack could occur as a result of fire.
However, the system is capable of being isolated in case of any fire, and burn itself out by cutting the oxygen to the fire.
elechJc-l el";)m J (4) Qualifications of Fire Barriers-The walls common with th "H p r -~'
cpg hr: r9 gr room (Rm 640), the SGTS A division 3 room (Rm 642), the b ceiling, and a section of the floor common to 6re area FS400 (Rm 543) below serve as fire barriers between adjacent fire areas and are of S h fire-resistive concrete construction. The remainder of the floor (not common to FS400),
the wall common with SLC Area and corridor B room 622 are not rated as they are internal to fire area F4201. A non-fire rated door provides access from corridor D (Rm 643).
l 9A4-293 Analysis
Rev.O ABWR oesien conretoecamemmer2 9A.4.1.6.34 Not Used 9A.4.1.6.35 FMCRD D/8 Panel Room (Rm No. 681)
(1) Fire Area-F7200 (2) Equipmenu See Table 9A.6-2 Safety-Related Provides Core Cooling Yes, D2 & D3 No (S) Radioactive Material Present-None that can be released as a result of fire.
(4) Qualifications of Fire Barriers-All walls, the ceiling, and the floor are of S h fire-resistive concrete construction. Access to room 681 is from stair well (Rm 329) and elevator (Rm 328) via 3 h rated fire-resistive doors. The room provides access to D/G B upper fan room (Rm 663) and to the upper !"/"I t
.r h: :=h=;;:: =Mr room (Rm 680) through 3 h rated fire-resistive dc, ors.
- c. P l k eiset r;c A glyf (5) Combustibles Present Ig t.
Fire Ioading Total Heat of Combustion (MJ) 2 Cable Tray 727 MJ/m2 NCLL (727 MJ/m maximum average) applies (6) Detection Provided-Class A supervised POC in the room and manual alarm pull stations at 1.4-E.0,1.7-C.0.
(7) Suppression Available:
Type Location / Actuation l
Standpipe and hose reel Col.1.4-E 0, and 1.7-C.0/ Manual ABC hand extinguishers Col.1.4-E.0, and 1.7-C.0/ Manual (8) Fire Protection Design Criteria Employed:
(a) The function is located in a separate fire-resistive enclosure.
(b) Fire detection and suppression capability is provided and accessible.
(c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers.
9A.4-306 Analysis
1
\\
Rev. o ABWR oeska comratDocumemmer2 (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System:
(a) Location of the manual hose suppression system external to the room (b) Provision of raised supports for the equipment (c) Refer to Section 3.4 " Water Level (Flood) Design", for the drain system.
(d) ANSI BSI.1 standpipe (rupture unlikely) i (12) Fire Containment or Inhibiting Methods Employed:
(a) The functions are located in a separate fire-resistive enclosure.
(b) The means of fire detection, suppression and alarming are provided and accessible.
(13) Remarks-None.
i g,,.3g ep; put-9A.4.1.6.4@^J: 7.......J l D ".:._...V g,
noom (Rm No. 640)
C I' d (1) Fire Area-F6200 4ft,/4 (2) Equipment See Table 9A.6 2 Safety-Related Provides Core Cooling No No (3) Radioactive Material Present-None that can be released as a result of fire.
(4) Qualifications of Fire Barriers-All walls and the floor are of S h fire-resistive concrete construction. A section of the ceiling is common to the FMCRD room (Rm 681) above and is of S h fire-resistive concrete construction. The remainder of the ceiling is internal to fire area F6200 and is not fire rated.
Access is provided from rooms 625 and 614 through S h fire-resistive doors.
(5) Combustibles Present Fire Loading Total Heat of Combustion (MJ) 2 2
Cable Tray 727 MJ/m NCLL (727 MJ/m maximum average) applies (6) Detection Prmided-Class A supervised POC in the room and manual alarm pull stations at 1.0-B.2 and 1.4-D.7.
Analysis 9A.4-314
Rev. 0 1
ABWR oeska canreroecumentmaz (10) Consequences of Fire Suppression-Suppression exdnguishes the fire. Refer to Section 3.4 " Water Level (Flood) Design", for the drain system.
(11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System:
(a) Location of the manual hose suppression system external to the room i
(b) Provision of raised supports for the equipment (c) Refer to Secdon 3.4 " Water Level (Flood) Design", for the drain system.
(d) ANSI BSI.1 standpipe (rupture unlikely)
(12) Fire Containment or Inhibidng Methods Employed:
(a) The functions are located in a separate fire-resistive enclosure.
(b) The means offire detection, suppression and alarming are provided and accessible.
(13) Remarks-None.
Room Rm No. 680) cpl 9A.4.1.6.47 L';;:::"?!M M::' E ^.:x; : rd ": 1c+(dca1 6[t, j9(,
(1) Fire Area-F6400 Oc (2) Equipment: See Table 9A.6-2 Safety-Related Provides Core Cooling No No (3) Radioactive Material Present-None.
(4) Qualifications of Fire Barrien-The walls in common with the FMCRD room (Rm 681), the SBGT filter train room (Rm 642), corridor B room (Rm 643),
both exterior walls and the ceiling are of 3 h fire-resistive concrete construction. The floor is common to room 640 below and is not fire rated.
Access to room 680 is provided from the FMCRD room via a 3 h fire-resistive door and directly from room 640 below via a stairwell.
t l
i l
9 i
Analysis SA,4 320
R3:10 ABWR outen cantoocamaatmerz Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued)
Leestion Loeotion item Elect Elev.
Nwnber Alphe system Room No.
MPL %
Div.I Laostion Coord.
Coord.
Desertption Drawing No.
2466 P54-F0038 2
23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640
- N
~
2467 P54-F0128 2
23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640
[
2468 P54-F203 N
23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640 N
2449 P54-PIS0015 2
23500 1.8 A.6 PRESS IND SWITCH 107E5128/0 640 2470 P54.PT004 N
23500 1.8 A.6 PRESS TRANSMITTER 107E5128/0 640 d) 2471 P63-1A N
23 1.5 A5 HEAT EX GER 55-812 247 P63-800 N
23500 1.2 A.5 HWHH EXCHANG 1002b58 640
)
9 1
(B) u 2473
-8002 23500 1.2
- 7tl.)
CKUP H EXCH 1 5-812 640.
W 2474 H22-P0448*
2 23500 1.7 B.8 CAMS GAS CYL RACK B 107E5139/1 640 p
- - =
n er 1.3 a.s
.:c;;; FcaF e, imm =
x.
N fMM W e;;;
N 2-i.3 5.7
.i.;;; FOMT i IGG;;;;ius e-2477 H23-P029' N
23600 1.9 C.3 MULTIPLEXER 640 2478 H23-P030' N
23500 1.9 C.5 MULTIPLEXER 640 2479 H23-P031' N
23500 1.9 C.7 MULTIPLEXER 640 2480 D11-F053 N
23500 2.4 C.9 SOLENOID VALVE 107E6071/0 641 2481 D11-F054 N
23500 2.4 C.9 SOLENOID VALVE 107E6071/0 641 2482 D11-RE002A N
23500 2.5 C.1 SBGTS ION CHAMBER 107E6071/0 641 2483 D11-RE0028 N
23500 2.5 C.1 SSGTS ION CHAMBER 107E6071/0 641 2464 H22-PO43B 2
23500 2.2 C.5 SBGT INSTR RACK 10Q273-285 641 2485 P54-DPS003 N
23500 2.0 C.5 D(FF PRESS SWITCH 107E5128/0 641 2486 T22-80018 2
23500 2.2 C.7 DRYER HEATER S 107E6046/1 641 2487 T22 C0018 2
23500 2.2 C.1 EXHAUST FAN B 107E6046/1 641 2488 T22.C002B 3
13500 2.2 C.7 COOUNG FAN B 107E6046/1 641 2489 T22-C0038' 2
23500 2.2 C.6 PREHTR & FAN B - FLTR 107E6046/1 641 i
2490 T22-C0048*
2 23500 2.2 C.6 AFTRHTR & FAN B FLTR 107E6046/1 641 2491 T22-Do03B 2
23500 2.2 C.6 PRE HEPA FILTER B 107E6046/1 641 2492 T22-0004B 2
23500 2.2 C.2 POST HEPA FILTER B 107E6046/1 641 2493 T22-00028*
2 23500 2.2 C.7 PRE FILTER TRAIN 107E6046/1 641 2494 T22-OP10038 2
23500 2.2 C.7 DIFF PRESS INDICATOR 107E6046/1 641 2495 T22-OP1007B 2
23500 2.2 C.7 DIFF PRESS INDICATOR 107E6046/1 641 2496 T22-OP1008B 2
23500 2.2 C.6 DIFF PRESS INDICATOR 107E6046/1 641 2497 T22-DP1012B 2
23500 2.2 C.6 DIFF PRESS INDICATOR 107E6046/1 641 2498 T22-DP10178 2
23500 2.2 C.2 DIFF PRESS INDICATOR 107E6046/1 641 Fire Hazard Analysis Database 9A.6 82
t insert "B" for Table 9A.6-2, Page 9A.6-82 gg Renlacement of Electrical Equipment for IIWil Equinmment 2471 R23 P/C ENI10A NI 23500 1.5 A.5 P/C ENI10A - LO VOLT SWTGR 107E5072/0 640 2472 R23 P/C ENI10B N2 23500 1.2 A.5 P/C ENI10B - LO VOLT SWTGR 107E5072/0 640 2473 R23 P/C EN110C N3 23500 1.2 A.2 P/C EN110C - LO VOLT SWTGR 107E5072/0 640 2475 R24 MCC ENI10A N1 23500 1.3 B.5 MCC ENI10A - R/B 107E5072/0 640 2475a R24 MCC ENI10B N2 23500 1.3 B.6 MCC ENI10B - R/B 107E5072/0 640 2476 R24 MCC ENI10C N3 23500 1.3 B.7 MCC ENIIOC - R/B 107E5072/0 640 x
1r o
1 bR 7
A e
g C9 %
H%
h
Rev. 0 ABWR omier canetcumanmu2
/
Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued)
Loestion Loee6en item Elect Elev.
Number Alpha System Room No.
MPt. No Div.
Loostion Coord.
Coord.
Desertp6en Drew 6ng No.
2674 D23-F197A 1
26000 5.3 D.9 SO VALVE 107E5130/1 000 2675 H22-P053A*
1 26900 5.3 0.9 D23, CAMS RACK A 107E5139/1 659 2676 H22-P054A' 1
26900 5.2 D.7 D23, CAMS CAUB RACK A 107E5130/1 659 2677 U41-0113 1
27200 5.3 D.7 CAMS (A) ROOM HVH 107E5109/0 656 2678 U41-C204B 2
27600 1.4 E.8 DG(B) HVAC SUPP FAN 8 107E5189/D 663 2679 U41 C204F 2
27600 1.2 E.8 DG(B) HVAC SUPP FAN F 107E5180/0 683 2680 U41 TE056 2
27600 1.4 EJ TEMP ELEMENT 107E5180/0 663 2681 P25-F022B 2
27600 1.2 F.3 TCV; DG 8 RM CLG 107E5182/0 683 2682 U41 C207C 3
27600 6.8 E.8 DG(C) HVAC SUPP FAN C 107E5180/0 673 2683 U41-C207G 3
27600 6.5 E.8 DG(C) HVAC SUPP FAN G 107E5100/0 U3 2684 U41 TE060 3
27600 6.7 E.8 TEMP ELEMENT 107E5189/0 673 2685 P25-F022C 3
27600 6.7 F.3 TCV: DG C RM CLG 107E5182/0 673 g'~2666
-F010 N
27000
.5 A.
TEMP NTROLV 10025 12 600 j
2 P63-27 1.5 A.5 P CONTRO ALVE 1
5-f12 MO, 688
-PT002 N
1.5 A.5 TEMPERAT E 10Q255-81.
Geo TRANS R
Sgt. 'S g
9(,h Ad89 P63- 005 27600 1.5 5
TE ELEMENT 1
55-812 680 dff(f
{ 2690 )43 TE007 / N 27 1.5 A.5 MP ELE 10Q255-812 w
2001 H21 P009-01 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2692 H21-P009-03 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2693 H21-P009-05 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2694 H21-P009-07 N
27600 1.5 D.0 REMOTE COMM CA8 NET 103E1167 681 (C11) 2695 H21-P009-09 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2696 H21 P009-11 N
27000 1.5 D.0 REMOTE COMM CA8 NET 103E1167 681 (C11) 2697 H21-P009-13 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2898 H21-P009-15 N
27600 1.5 D.0 REMC TE COMM CABNET 103E1167 681 (C11) 2699 H21-P009-17 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2700 H21-P009-19 N
27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 9A.&89 Fire Hatsid Analysis Database
Rev.0 ABWR omina cueetonsmanmu2 i
14.2.12.1.30 Not Used 14.2.12.1.31 Hot Water Heating System Preoperational Test (1) Purpose Vedfy the ability of the Hot Water Heating System (HWHS) to provide hot water to the appropriate HVAC systems and the operation of HWH pump, heat exchanger, surge tank and chemical addition tank.
(2) Prerequisites The construction tests have been completed, and the SCG has reviewed the test procedure and approved the initiation of testing. Electrical power, SA c, p t System, TCW System, Heat Steam System, T- - - r " _" F g Oc-shg W=:
o D
0.a.;, HVAC System, HNCW System and other required interfacing systems fg 7
shall be avadable, as needed, to support the specified testing. Additionally, a temporary strainer shall be installed at the suction side of the HWH pump.
(3) General Test Methods and Acceptance Cdteda Performance shall be observed and recorded dudng a series ofindividual
(
component and integrated system tests. These tests shall demonstrate that the HWHS operates properly as spect6ed in appropdate HWHS design specification and manufacturer's technicalinstruction manual through the following testing:
(a) Proper opention ofinstrumentation and system controls in all combinations oflogic and instrument channel trip (b) Verification ofvarious component alarms, for correct system respon
- to process variable, and provides r.larms at the prescribed value (c) Proper operation of system valves, including open/ closure cycling and posidon indicator verificadon, if applicable (d) Proper operating conditions (flow, vibration, bearing temperature) of the HWH pumps dudng condnuous pump run test (e) Acceptable pump NPSH under the mostlimidag design flow conditions.
(f)
Proper operating conditions and system performance capability during the following operation mode tests:
(i)
Plant normal operation mode (ii) Plant shutdown and inspection mode (g) Proper pump motor start sequence and actuation of protective devices SpecrRc Information to be included in Final Safety Analysis Repons N.2-00
)
i nov. o Aswg seals.cssents.ssessper
-(.
(h) Proper operation of interlock functions, including operation of all components subject to interlockmg (e.g., HWHS pump trip on low surge tanklevel
, r : n;2 d"""~.J- ' 231,.n-L- p.d cp i and system water temperature control, etc.)
bhlfG (i)
Proper operation of pernussive, prohibit, and bypass functions (j)
Proper operation of system surge tank and chemical addition tank and their assocuted functions dunng system operation mode tests l
14.2.12.1.32 HVAC Emergency Cooling Water System Prooperational Tect (1) Purpose i
To verify the ability of the HVAC Emergency Coohng Water (HECW) System to supply the design quantities of chilled water at the specified temperatures to the various cooling coils of the HVAC systems serving rooms and areas containing es.cntial systems and equipment.
(2) Prerequisites The construction tesu have been successfully completed, and the SCG has reviewed the test procedure and approved the initiation of testmg. Normal and aimliary electrical power, IA, MUWP, RCW, applicable HVAC System cooling coils, and other required system interfaces shall be available, as needed, to support the specified system testing.
(3) GeneralTestMethods and Acceptance Criteria Performance shall be observed and recorded during a series ofindividual component and integrated system tests.These tests shall demonstrate that the HECW System and its aimiliary equipment operate properly as specified in Subsections 9.2.13 and 7.3.1.1.9 and applicable HECW System design specification through the following testir.g:
(a) Proper operation ofinstrumentation and system control functions including flow switch, surge tank level controller, and chilled water temperature controller (b) Verification of various component alarms, for correct alarm actuation and reset, alarm set value, alarm indication and operating logic (c) Proper operation of system motor operated and air operated valves, including operability and position indication verifications,if applicable (d) Proper operation of HECW pumps and motors during continuous run tests 14.241 Specific Information to be included in Final Seiety Analyeie Reporn
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Figure 1.2-10 REACTOR BUILDING, ARRANGEMENT PLAN AT ELEVATION 23500mm 1
l ABWR DCD/Tler 2 nev. o it-ts 3
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- Equipment receives BCW in this mode.
- uo;e on The HECW System is manually initiated. ((*,.'l '.,'\\ j
- -l AREA HVAC SYSTEM 4
- 1. DIVISION A IS POWERED FROM CLASS 1E DIVISION 1.
- 1. The basic configuration for the HECW
- 1. Visual inspections of the as-built system 1.
- 2. The ASME Code components of the
- 2. A hydrostatic test win be conducted on
- 2. The results of the hydrostatic test of the It HECW System retain their integrity under those Code components of the HECW ASME Code components of the HECW internal pressures that will be System required to be hydrostatically System conform with the requirements in l
- 3. Each HEWC System refrigerator unit has a 3. Type tests will be conducted on an as-
- 3. Each HEWC System refngerator unit has a capacity of not less than 2.43 GJ/h.
- 4. Tests win be conducted on each as-built in Divisions B and C, the refrigerator unit Min Divisions B and C, any refrigerator unit on standby automatically starts if any of JCW System refrigerator unitin on standby automatically starts upon the other refrigerator units in Divisions -
- 5 1
- a. Tests will be performed on the HECW a.
- b. Inspectens of the as-built Class 1E
- b. In the HECW System, physical divisions and non-Class 1E equipment.
- 7 ;[ M gf " D(*V N L *.D;,Is; A,
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- 88 uCC INST ANI tNTE RRUP-NE t 0 R/B ESSENilAL ELEC isON COUNTER uC AS NO T NE E D EOUtP ROOutBI INCHING NEED VENIIL A TOR 186 RUNNtNC ON HECW PuuPS IC00tC & COO 1FI E ACH SUFFIX 8 AND E CHANCES TO C AND F OPE R A llON NO T NE E D HOwEVER. HECW PUuP 18 AND El 15 uCR VENitL A TOR t Al VALVE SE A TING OPEN StOE uCC AND HECW PuuP (C AND FB IS uCR VENiiL ATOR 181 FORu CLOSE D StDE HECw PUuP E81
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- (o Capacity (Refrigerator) 2.51 GJ 3
- r..:.:..^. in.= ;;m;..:..;; _ x? Divisions B and C e two pa<.lle l pump-refrigerator units.
- 7s5D 5.So 32 U41 C805A 17190 5.20 K.5 EM ElfC (A) EXH FAN A 10tE5188/O 813 2s& A 217 218 U41 C806E 1
- n System configurations for each division are illustrated in Figure 19D.S.1,19D.S 2 and SD.S.S.
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nev. 0 oseios casentooemenmaz ABWR (2) Diesel generator selection shall include prudent component design with dust tight enclosures. Construction guidelines shall include provmons for mimmning accumulation of dust and dirt into equipment. These shall be in accordance with recommendations 2.a. 2.b 2.d and 5 of NUREG/CR 0660 (Subsection 9.5.6.5).
(S) The diesel generator operating procedure shall include provisions to avoid as much as possible or otherwise restrict the no4oad or low load operation of the engine / generator for prolonged periods of time; or operate the engine at nearly full-load following every no-load or low 4oad (20% or less) operation lasting for a period of 30 minutes or more (Subsection 8.5.1.1.8).
9.5.13.9 Applicant Rre Protection Program The following areas are out of the ABWR Standard Plant design scope for the fire protection program, and shall be included in the COL applicant's fire protection program:
(1) Main transformer (2) Equipment ennylock (3) Fire protection pumphouse (4) Ultimate heat sink The COL applicant's fire protection program shall comply with the SRP Section 9.5.1, with ability to bring the plant to safe shutdown condition following a complete fire burnout of a fire area / division without a need for recovery (Subsection 9.5.1).
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e COL apphcant shall provide pressure calesbnons and confirm capability durin preoperationaltesting of the smoke control mode of the HVAC systems as described in Sub=*ction 9.5.1.1.6.
M 9.5.13.11 P:ent Security Systems Crtteria c (=.3 YYN The COL applicant's design of the security system (Subsection 9.5.2) shallinclude an evaluation ofits impact on plant operation, testing, and maintenance. This evaluation shall assure that the security restrictions for access to equipa nt and plant regions is compatible with required operator actions during all oper.r.4 and emergency modes s.s.72 other Auxinery synoms
Cev. O ABWR D**** Co***I D"*=**2 Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued)
Looselon lasselen hem Elset Elev.
hainer Alphe system Room No.
MPL h Div.
Laeselen Comed.
Coord.
Deseripelen Drawin8 No.
2331 U41-52028 1
27200 6.8 A.8 COOL COIL ELEC EQ(A) 107E5180/0 812 2332 U41-5202A 1
27200 6.5 A.8 COOL COIL ELEC EQ (A) 107E5180/0 812 2333 U41-F005A 1
27200 6.4 A.5 MO VALVE 1m5180/0 613 2334 P54-A001A 1
23500 8.2 B.2 N2 STORAGE BOTTLE 107E5128/0 813 2335 P54-A001C 1
23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 813 2335 P54-A001E 1
23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 813 2337 P54 A001G 1
23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 813 2338 P54-A001J 1
23000 8.2 B.2 N2 STORAGE SC'TTLE 107E5128/0 813 2330 PE*-A001L 1
23500 8.2 B.2 N2 STORAGE BOTTLE 107E5128/0 813 2340 P54-A001N 1
23500 8.2 B.2 N2 STORAGE BOTTLE 107E512Al0 813 2341 P54-A0010 1
23500 8.2 B.2 N2 STORAGE BOTTLE IM5128/0 813 2342 P54-A0015 1
23500 8.2 B.2 N2 STORAGE BOTTLE 1m812Sl0 813 2343 P54-A001U 1
23500 8.2 B.2 N2 STORAGE BOTTLE 107E5128/O 813 2344 P54-F003A 1
23500 8.2 B.2 MO GLOBE VALVE 107E5128/0 813 2345 P54-F012A 1
23500 8.2 B.2 MO GLOBE VALVE 107E5120WO 813 2346 P54-PIS001A 1
23500 8.2 B.2 PRESS IND SWITCH 1m612E/0 613 I
'C b 3 2347 U41-C202A 1
23500 6.4 B.2 DGW HVAC EXH FAN A 107E5158/0 813 Ob 4b 2348 U41.C202E 1
23500 6.4 B.5 DG(A) HVAC EXH FAN E 107E5188vo 813 A dJ - -->
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umt 2350 R24 MCC A310 N
23500 6.3 C.0 MCC A310 - R/B 107E5072/0 813 2351 R43 C201 A*
1 23500 6.6 B.8 DG AIR COMPRESSOR A SSAR FIG 9.54 813 2352 R43-C202A' 1
23500 6.9 B.8 DG AIR COMPRESSOR A SSAR FIG 9.5-8 813 2353 D21-RE007 N
23000 5.2 B.0 AREA RAD DETECTOR 299X701-171/0 815 2354 T31-SSA051 N
23500 5.3 C.3 SELECT SWITCH 107E8043/0 815 2355 T31-SSA053 N
23500 5.3 C.3 SELECT SWITCH 107E8043/0 815 2356 T31-T1051 N
23500 5.3 C.4 TEMP INDICATOR 107E8043/0 815 2357 T31-T1053 N
23500 5.3 C.3 TEMP INDICATOR 107E8043/0 815 2350 T31-TTQ51 N
23000 5.3 C.3 TEMP TMANSMITTER 107E8043/0 815 2350 T31-TT053 N
23000 5.3 C.3 TEMP TRANSMITTER 107E8043/0 815 2300 U41-C103 N
23500 5.4 C.1 PCV PURGE SUPPLY FAN 107E5188/0 815 2381 U41.P004C 3
29000 5.8 B.8 MO VALVE 107E510Bl0 816 I
2382 U41-F101C 3
29000 5.8 B.8 MO VALVE 107E5180/0
$10 2383 T31-F731 1
23500 5.8 C.8 SO VALVE 107E8043/0 816 2384 T31-FT054 N
23500 5.8 C.8 PRESSURETRANSMfTTER 107E8043/0 816 Fire NatardAnalysis Database SA.0-78
nov. o ABWR onnee coneetoecomeetmer2 Table 9A.6 2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued)
Leeseen Leestion hem West Elev.
Number Alpha Byeena Room No.
MPL N.
Div.
Leestion Coord.
Coeral Description Drawing No.
2432 C41. TIS 006 N
23500 2.3 D.6 TEMP SWITCH 107E6016/0 622 2433 U41-C104 N
23500 2.3 E.7 PCV PURGE EXHAUST 107E5180/0 823 FAN 2434 R43-LS306B' 2
23500 1.3 E.9 LEVEL SWITCH SSAR FIG 9.H 624 2435 U41-8204B 2
27200 1.2 F.2 COOL COIL.ELEC EQ (B) 107E5188/0 083 2438 U41-8204F 2
27200 1.5 F.2 COOL COILELEC EQ (B) 107E5188/0 863 i
2437 U41-F005B 2
27200 1.6 F.3 MO VALVE 107E5180/0 083 g3 U41-C206F 2
23500 1.8 E.7 DG(B) HVAC EXH FAN F 107E5188/0
$25 A dd-,[2438 439 U41-F008B 2
23500 1.6 E.5 MO VALVE 107E5180/0 625 g
D 3erf 2440 R434201B' 2
23500 1.1 E.2 DG AIR COMPRESSOR B SSAR FIG 9.H $25 f
"I 2441 R4342028' 2
23500 1.3 E.2 DG AIR COMPRESSOR B SSAR FIG 9.5 8 425 2442 U41-C205B 2
23500 1J E.4 DG(B) HVAC EXH FAN B 107E5188/0 825 2443 R24 MCC B310 N
23500 1.8 D.5 MCC B310 - R/B 107E5072/0 825 2444 R43-A005C*
3 23500 6.5 F.8 FUEL OIL DAY TANK SSAR FIG 9.H 830 2448 R43-LS305C' 3
23500 S.7 E.9 LEVEL SWITCH SSAR FIG 9.5 6 632 2448 U41-8200G 3
27200 6.8 F.2 COOL Coll.ELEC EQ (C) 107E5188/0 573 l
2447 U41-8200C 3
27200 6.5 F.2 COOL ColLELEC EQ (C) 107E5188/0 673 2448 H22-P044A' 1
23500 6.3 F.1 CAMS GAS CYL RACK A 107E5130/1 833 Jeet U41-F005C 3
27200 6.4 F.3 MO VALVE 107E5188/0 873 2450 R24 MCC C310 N
23500 6.3 E.5 MCC C310 R/B 107E5072/0 833 L 3 e / t-2461 R43-C201C' 3
23500 6.6 E.2 DG AIR COMPRESSOR C SSAR FIG 9.5 8 833 "A"
2462 R43-C202C' 3
23500 6.6 E.4 DG AIR COMPRESSOR C SSAR FIG 9.H 833 f
2453 R10-C001E' N
23500 6.3 D.5 RIP ASD OUTPUT XFMR 838 2464 R10 C0018' N
23500 6.3 C.7 RIP ASD OUTPUT XFMR 838 2455 U41-01345 N
23500 5.5 E.4 ist ROOM FCU B 107E5100/0 830 2456 P54-A0018 2
23000 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2487 P54-A001D 2
23000 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2468 P54-A001F 2
23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2450 P54-A001H 2
23000 1.8 A.4 N2 STORAGE BOTTLE 107E5120/0 640 2400 P54-A001K 2
23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2461 P54-A001M 2
23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/O 640 2462 P54 A001P 2
23000 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2463 P54-A001R 2
23500 1.8 A.4 N2 STORAGE BOTTE 107E5128/0 640 2464 P54-A001T 2
23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2486 P54-A001V 2
2;500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 9A.6-81 Fire Metard Analysis Datsbene
N Insert "A" for Table 9A.6-2, Pages 9A.6-78 and -81 CP3 Addition of Smoke Removal Fan in R/B S/R Electrical Equipment IIVAC (Al (B) and (C) 2348a U41-C210A N
23500 6.4 B.5 SREE IIVAC Smoke Removal Fan A 107E5189/0 613 2438a U41-C211B N
23500 1.8 E.7 SREE IIVAC Smoke Removal Fan B 107E5189/0 625 2449a U41-C212C N
23500 6.3 E.5 SREE IIVAC Smoke Removal Fan C 107E5189/0 633 OO J
hh
~
>c mm i
I
Rev. O ABWR oesien conedseementria2 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room -
Control Building (Continued)
LOCADON LOCADON ITEM MPL ELECT ELEV.
NUMBER ALPHA SYSTEM ROOM NO.
NO.
DIV.
LOCATION COORD COOIW DESCRFnON DRAWING NO.
200 P25-D001G 3
12300 6.70 J.2 HECW REFRIGERATOR F 107E51824 534 201 P25-DFT007C 3 12300 5.70 J.6 DP XMTR (FLO CONT CE) 107E51824 534 202 P25-F005C 3
12300 6.90 J.2 TCV: MCR CLG 107E51824 534 203 P25-F012C 3
12300 5.70 J.4 PCV: HECW UNITS C/F 107E51824 534 204 P25-FIS003C 3
12300 6.00 J.2 FLOW IND SWITCH C 107E51824 534 205 P25-FiS003F 3
12300 6.70 J.2 FLOW IND SWITCH F 107E51824 534 f
206 P25-TE005C 3
12300 6.00 J.2 TEMP ELEM (UNIT CE) 107E5182A 534 I
207 U41 C623C 3
12300 6.20 J.1 MCR RECIRC SUPP FAN C 107E518tl0 534 208 U41.C623G 3
12300 6.20 J.1 MCR RECIRC SUPP FAN G 107E5188/0 534 209 H11-P001' N
12300 4.00 K.0 COMPUTER PANELS 501 210 P25-A002 N
12300 5.30 J.5 CHEMICAL FEED TANK 107E51824 593 211 P25-DFT007A 1 12300 5.30 J.2 DP XMTR (FLO CONT A) 107E51824 593 212 P25.TE006A 1
12300 5.30 J.2 TEMP ELIM (UNIT Al 107E51824 583 213 P25-C001A 1
17150 5.30 J.4 HECW PUMP A 107E51824 612 214 P25-D001A 1
17150 5.30 J.4 HECW REFRIGERATOR A 107E51824 612 215 P25-F012A 1
17150 5.50 J.2 PCV:HECW UNIT A 107E51824 612 216 P25-FIS003A 1
17150 5.30 J.2 FLOW IND SWITCH A 107E51824 612 l
217 U41-C805A 1
17150 5.20 K.5 EM ELEC(A)EXH FAN A 107E5188/0 613 218 U41 C605E 1
17150 5.20 K.6 EM ELEC (Al EXH FAN E 107E5188/0 613
- '."'Yh Y " 2 *I' + Q y,,,g.
g--
y.
g,y u.
220 U41-C622C 3
17150 5.70 K.5 MCR HVAC EXH FAN C 107E5180/0 614 221 U41-C622G 3
17150 5.70 K.6 MCR HVAC EXH FAN G 107E518elo 614 222 V41-DP1106C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5180/0 614 l
C P3 ;
223 U41-DP1107C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 C/ f4b 224 U41-DP1100C 3 17150 5.70 K.5 DIFF PRES $ 1NDICATOR 107E5189/0 614 225 U41-OP1109C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5188/0 614 226 U41-F009C 3
17150 5.70 K.5 MO VALVE 107E51894 614 227 U41-F000G 3
17150 5.70 K.6 MO VALVE 107E5189/0 614 228 U41-F010C 3
17150 5.70 K.5 MO VALVE 107E5189/0 614 229 U41-F010G 3
17150 5.70 K.6 MO VALVE 107E5199/0 614 pele t ect ") N w%
a A :n )la u
'9*0 "M
MO '/n'/
'"M" T
"10
-y 230 00: T0'20 CIb 231 U41. POT 105C 3 17150 5.70 K.5 POSITION TRANSMITTER 107E5189/0 614
{
232 U41 POT 105G 3 17150 5.70 K.6 POSITION TRANSMITTER 107E5189/0 614 233 U418601C 3
17150 6.80 J.3 MCR COOUNG Coll 107E5189/0 615 Fire HatardAnalysis Database 9A.6-102
Rev. 0 ABWR kuen coneetoecanmaamu2 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sarted by Room -
Control Building (Continued)
LOCATION LOCATION ITEM A8PL EECT ELEV.
NUMBER ALPHA SWEM ROOM NO.
NO.
DIV.
LOCATION COORD COORD DESCRrHON ORAWWWG NO.
288 U41-F0098 2
17150 1.80 J.1 MO VALVE 107E51584 821 289 U41-F000F 2
17150 1.00 J.1 MO VALVE 107E518M) 821 270 U41-F011B 2
1715C 1.50 J.1 MO VALVE 107E51884 821 271 U41-ME1048 2 17150 1.50 J.1 MOfSTURE EMMENT 107E51894 821 272 U41-TE1038 2
17150 1.50 J.1 TEMP EMMENT 107E51894 821 273 P25 C001B 2
17150 2.80 J.4 HECW PUMP 8 107E51824 823 274 P25-C001E 2
17150 2.80 J.8 HECW PUMP E 107E51824 823 275 P25-D0018 2
17150 2.00 J.4 HECW REFRIGERATOR 8 107E51824 823 276 P25-D001E 2
17150 2.80 JJ HECW REFRIGERATOR E 107E51824 823 277 P25-DPT0078 2 17150 2.30 J.2 DP XMTR(FLO CONT 8E) 107E51824 823 278 P25-F0128 2
17150 2.F0 J.2 PCV: HECW UNITS 84 107E5182A 423 279 P25-FIS0038 2
17150 2.80 J.4 FLOWIND SWITCH 8 107E51824 823 200 P25-FIS003E 2
17150 2.80 J.8 FLOW IND SWITCH E 107E5182A 823 281 P25-TE0058 2
17150 2.80 J.4 TEMP ELEM (UNIT BE) 107E51824 623 282 P25-F0168 2
17150 1.70 K.8 TCV:C4 ELEC RM B 107E51824 824 283 U41-88048
't 17150 1.70 K.8 ESS EQUIP RM COOL COIL 107E51884 824 284 U418804F 2
17150 1.70 K.8 ESS EQUIP RM COOL Coll 107E518mm 824 285 U41-C0068 2
17150 1.80 K.5 EM ELEC (8) SUPP FAN 8 107E518A/O 824 286 U41-C606F 2
17150 1.00 K.6 EM ELEC (B) SUPP FAN F 107E5189/0 624 287 U41-DP11118 2 17150 1.80 K.8 DIFF PRESS INDICATOR 107E5180/0 624 /
288 U41-F1048 2
17150 1.80 K.8 MO VALVE 107E5180M 824 289 U41 TE110B 2
17150 1.80 L1 TEMP EMMENT 107E5189/0 624 (
290 U41 TE1128 2
17150 2.00 K.6 TEMP ELEMENT 107E5189/0 624 291 U41-C607B 2
17150 2.20 K.5 EM ELEC (B) EXH FAN 8 107E51894 825 cP3 292 U41 C807F 2
17150 2.20 K.6 EM ELEC (B) EXH FAN F 107E5189/0 625 293 Z: """""
2
" " "0 2.20 1"
"30 vE'M "On"CT C i 3 294 U41.TE1138 2
17150 2.20 K.5 TEMP ELEMENT 107E5189/0 625
.o b 295 U41-C602B 2
17150 2.70 K.5 MCR HVAC EXH FAN B 107E5189M 626 b
j 296 U41-C602F 2
17150 2.70 K.6 MCR HVAC EXH FAN F 107E5189/0 626 e3 i
297 U41-F0108 2
17150 2.70 K.5 MO VALVE 107E5189/0 626 g
298
:-
2
=
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meer e 299 U41. POT 1058 2 17150 2.70 K.5 POSITION TRANSMITTER 107E51894 626 f
[
300 U41-C809C 3
17150 8.20 K.5 EM ELEC (C) EXH FAN C 107E5189/0 631
^
301 U41 C600G 3
17150 6.20 K.6 EM ELEC (C) EXH FAN G 107E51894 631 fire Hazard Analysis Database 9A.6-104
m I
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Ry. 0 ABWR outen coneetoocanuttrier2 i
(
Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room -
$)
Control Building (Continued) 3 "o
3, a*
LOCAnON LOCAnON ITEM MPL ELECT ELEY.
NUMBER ALPHA SYSTEM ROOM 1
NO.
NO.
DIV.
LOCATION COORD COOfC DESCRFTION DRAWING NO.
w QC, 302 Wi ";%C
;Z
".2^
K.:
'#0 i'^1t';
- 7::;-,
=;
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303 U41-OP1111C 3 17150 5.80 K.8 DIFF PRESS INDICATOR 107E5189/0 663 cP3 304 U41-F104C 3
17150 5.80 K.8 MO VALVE 107E5189/D 663 t
I 305 U41-TE110C 3
17150 5.90 L1 TEMP ELEMENT 107E5189/0 663 1
CP3 l
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i 9A.6-105 1
Fire Haznid Analysis Ostabase i
i Rev.o l
ABWR oesion controloocumenvrier2 l
l Table of Contents (Continued) 9A.4.2.3.23 Elevator (Rm No. 337).
. 9A.4-412 9A.4.2.4 Floor Four El 7900mm.
. 9A.4-413
. 9A.4-413 9A.4.2.4.1 Control Room Complex [.
l l
9A.4.2.5 Floor Five El 12300mm..
. 9A.4-417 y9A.4.2.5.1 Control Room HVAC "C" Exhaust Duct Chase (Rm No. 522)...
. 9A.4-417 9A.4.2.5.2 HVAC "A" Supply (Rm Nos. 511,512 and 513)...
.. 9A.4 419 9A.4.2.5.3 HVAC "C" Supply (Rm Nos. 531,532 and 533).
. 9A.4-421 9A.4.2.5.4 Stairwell Landing (Rm No. 505)..
.. 9A.4-423 9A.4.2.5.5 Chiller Unit "C" (Rm No. 534).
. 9A.4-425 9A.4.2.5.6 Recirc Internal Pump MG Sets and Control Panels (Rm Nos. 501,502,503 and 504).
. 9A.4 427 9A.4.2.5.7 Computer Room (Rm No. 591)...
. 9A.4-429 9A.4.2.5.8 Passageway (Rm No. 521).
... 9A.4-431 9A.4.2.5.9 Not Used...
. 9A.4-433
. 9A.4-433 9A.4.2.5.10 Passageway (Rm No. 592).
..d......
9A.4.2.5.11 Passageways (Rm No. 59
. 9A.4-435 9A.4.2.5.12 Control Room HVAC " ", Exhaust Duct Chase (Rm No. 595).
. 9A.4-437 CP4 9A.4.2.5.13 Passageway (Rm No. 506)..
. 9A.4-438
(, l (,f 9 (.,
. 9A.4-440 9A.4.2.6 Floor Six El 17150mm..
9A.4.2.6.1 Control Room HVAC Supply 'B" (Rm No. 621).
. 9A.4-440 9A.4.2.6.2 Passageway and Room (Rm No. 622 and 662).
. 9A.4-442
- Y 9A.4.2.6.3 Chiller Unit "B" (Rm No. 623).
.. 9A.4-444 g.
9A.4.2.6.4 HVAC "B" Supply and Exhaust (Rm Nos. 624,625, 627,661, and 664)..
. 9A.4-446
- 7
- " T 4 9A.4.2.6.5 HVAC "A" Intake Duct and Exhaustp Nos. 613,
- I M 1 l
617,618 and 619)....
. 9A 4-448
)
N N) y 9A.4.2.6.6 Control Room HVAC Exhaust " " (Rm Nos. 62p,2.... 9A.4-450
- SCO,029, and 000)P W ' hal
" 'd 4
9A.4.2.6.7 Chiller Unit "A" (Rm No. 612).
... 9A.4-452 9A.4.2.6.8 Control Room HVAC Supply "C" (Rm No. 615).
. 9A.4-454 4
9A.4.2.6.9 Passageway and Room (Rm Nos. 611 and 652)....
. 9A.4-456 NM 9A.4.2.6.10 Control Room HVAC Exhaust "B" (Rm Nos. 6g
[p g,y ""-.1 f
ElS r.d SM).%..
...$.5..
. 9A.4-458 A s +.S.
9A.4.2.6.11 HVAC "C" Intake Duct and Exhaust (Rm Nos. 631, V
632,633,634,651, and 653).
. 9A.4-460 9A.4.3 Turbine Building......
..... 9A.4-462 l
9A.4.3.1 Floor One El 5.3m..
. 9A.4-462 9A.4.3.1.1 Floor One (Except Fire Areas FT 1501-FT1503).
... 9A.4-462 9A.4.3.1.2 Air Compressors and Dryer Area (Rm No.111)..
. 9A.4-466 9A.4.3.1.3 Stair Tower #1 (Rm No. I14).
... 9A.4-467 9A.4.3.1.4 Stair Tower #2 (Rm No.122).
. 9A.4-469 9A.4.3.2 Floor Two El 12.Sm..
.. 9A.4-470 9A.4.3.2.1 Floor Two (Except Fire Areas FT 1501, FT 2500-i FT2505)......
. 9A.4-470 Table of Contents 9A.0-viii
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NOTES:
- 1. THE INBOARD ISOLATION VA!.VE IS POWERED FROM CLASS 1E DIVISION 11, AND THE OUTBOARD ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION 1.
Figure 2.11.5 HVAC Normal Cooling Water System 2.11.5 2 HVAC Normal Coohng Water System
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Ry. O ABWR oneien ceneroecamentmer s 1
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BSF-Basement,3rd floor NOTES:
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- 3. "*" Denotes watertight door.
l Control and Instrumentation Cables:
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l Appendix A-5 l
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l Rev. O ABWR oesien coneetoecamenvriaa I
List of Tables (Continued)
Table 9.4-4g HVAC System Component Descriptions-Non-Safety-Related Fans.......... 9.4 47 Table 9.4-4h HVAC System Component Descriptions-Non-Safety-Related Filters......... 9.4-47 6h 46 Table 9.4-41 HVAC System Component Descriptions-Non-Safety-Related 3
'ih h Cci! U ni ts...................................................................
l Are Hudt.'e Table 9.4-5 Turbine Building and Electrical Building HVAC System-Non-Safety-Related Equip m e n t Lis t................................................................................ 9.4-49 Table 9.5-1 Normal and/or Standby Lighting (Non-Class IE AC Power Supply)........... 9.5-77 Table 9.5 2 Lighting and Power Sources..................................................................... 9.5 78 Table 9.5 3 Stan dby Ligh tin g....................... -.................................................................. 9.5-78 Table 9.5-4 DC Emergency Ughting (Class IE DC Power Supply)................................ 9.5-79 Table 9.5-5 Summary of Automatic Fire Suppression Systems......................................... 9.5-80 Table 9A.2-1 Co re Coolin g............................................................................................. 9A.2-8 j
i Table 9A.51 Redundant Instrumentation or Equipment in Same Fire Area.................. 9A.518 1
Table 9A.5-2 Summary of the Reactor Building Special Cases........................................... 9A.5-19 Table 9A.61 Fire Hazard Analysis, Equipment Data Base - Sorted by MPL Number (Superceded by Table 9A.6-2)....................................................................... 9A.S5 Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building...........................................................................................................9A.6-6 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control B uildi n g................................................................................................... 9A.
1 Table 9A.6-4 Fire Hazard Analysis Equipment Database-Sorted by Room-Turbine B uildin g........................................................................................... 9A.6-1 Table 9B-1 Estimated Fire Severity for Offices and Light Commercial Occupancies........ 9B-12
. 9B 13 Table 9B.2 Fire Severity Expected by Occupancy..........................................
Table 9B-3 Cable Type and Configuration for UI Tests...................................................... 9B-14
. 9B-14 Table 9B-4 Summary of Burning Rate Calculations.........................................
9.0-v/vs List of Tables
--_ ~
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Table 9.4-41 HVAC Syst omponent Descriptions--Non-Safety-Related Fa
/O:'y..:g(Response to Question 430.243)
- A.- h d b g Non-Safety-Relatg:n C:M Units Quanttty Capacity (MJ/h)
Main Steam Tunnel 2
528.02 1
83.74 Refueling Machine Control Room ISI Rcom
/l 54.43 2
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- The COL applicant shall supply equipment lists for the Servics Building HVAC and the Radwaste Building HVAC System. See Subsection 9.4.10.1 for the Service Building, and 9.4.10.2 for the Radwaste Building.
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I nav. O ABWR oesian contraroccamenvrier2 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room -
Control Building (Continued)
LOCATION LOCATION ITEM MPL ELECT ELEV.
NUMBER ALPHA SYSTEM ROOM NO.
NO.
DIV.
LOCATION COORD.
COORD.
DESCRIPTION DRAWING NO.
200 P25-D001G 3
12300 6.70 J.2 HECW REFRIGERATOR F 107E5182/0 534 j
201 P25-DPT007C 3 12300 5.70 J.6 DP XMTR (FLO CONT C/F) 107E5182/0 534 202 P25-F005C 3
12300 6.90 J.2 TCV: MCR CLG 107E5182/0 534 203 P25-F012C 3
12300 5.70 J.4 PCV: HECW UNITS C/F 107E5182/0 534 204 P25-FIS003C 3
12300 6.00 J.2 FLOW IND SWITCH C 107E5182/0 534 205 P25-FIS003F 3
12300 6.70 J.2 FLOW IND SWITCH F 107E5182/0 534 206 P25-TE005C 3
12300 6.00 J.2 TEMP ELEM (UNIT C/F) 107E5182/0 534 207 U41-C623C 3
12300 6.20 J.1 MCR RECIRC SUPP FAN C 107E5189/0 534 208 U41-C623G 3
12300 6.20 J.1 MCR RECIRC SUPP FAN G 107E5189/0 534 209 H11 P001*
N 12300 4.00 K.0 COMPUTER PANELS 591 210 P25-A002 N
12300 5.30 J.5 CHEMICAL FEED TANK 107E5182/0 593 i
211 P25-DPT007A 1 12300 5.30 J.2 DP XMTR (FLO CONT A) 107E5182/0 593 212 P25-TE005A 1
12300 5.30 J.2 TEMP ELEM (UNIT A) 107E5182/0 593 213 P25-C001A 1
17150 5.30 J.4 HECW PUMP A 107E5182/0 612 214 P25-D001 A.
1 17150 5.30 J.4 HECW REFRIGERATOR A 107E5182/0 612 215 P25-F012A 1
17150 5.50 J.2 PCV: HECW UNIT A 107E5182/0 612 216 P25-FiS003A 1
17150 5.30 J.2 FLOWIND SWITCH A 107E5182/0 612 (pg 217 U41-C605A 1
17150 5.20 K.5 EM ELEC (A) EXH FAN A 107E5189/0 613 6/qu 218 U41-C605E 1
17150 5.20 K.6 EM ELEC (A) EXH FAN E 107E5189/0 613 219 U41-F105A 1
17150 5.20 K.6 MO VALVE 107E5189/0 613 220 U41-C622C 3
17150 5.70 K.5 MCR HVAC EXH FAN C 107E5189/0 614 221 U41-C622G 3
17150 5.70 K.6 MCR HVAC EXH FAN G 107E5189/0 614 222 U41-DP1106C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 223 U41-DPl107C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 224 U41-DP1108C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 225 U41-DPl109C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 226 U41-F009C 3
17150 5.70 K.5 MOVALVE 107E5189/0 p
p
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227 U41-F009
.7 1 17150 5.70 K.6 MOVALVE 107E5189/0 SW"
228 U41-F010C 3
17150 5.70 K.5 MO VALVE 107E5189/0 614 1
l cP6 229 U41-F010 17150 5.70 K.6 MO VALVE 107E5189/0 614 230 U41-F012C 3
17150 5.70 K.5 MO VALVE 107E5189/0 614 231 U41-POT 105C 3 17150 5.70 K.5 POSITION TRANSMITTER 107E5189/0 614 232 U41-POT 105G 3 17150 5.70 K.6 POSITION TRANSMITTER 107E5189/0 614.,,
233 U41.B601C 3
17150 6.80 J.3 MCR COOUNG COLL 107E5189/0 615 9A.6102 Fire Hazard Analysis Database
4 skv. O ABWR Desien controlDocumenvrier2 Table 9A.6-3 Fire Hazard Analysis Equiprnent Data Base - Sorted by Room -
Control Building (Continued)
LOCATION LOCATION ITEM MPL ELECT ELEV.
NUMBER ALPHA SYSTEM ROOM NO.
NO.
DIV.
LOCATION COORD.
COORD.
DESCRIPTION DRAWING NO.
234 U41-8601E 3
17150 6.80 J.3 MCR COOLING COIL 107E5189/0 615 235 U41 B601G 3
17150 6.80 J.3 MCR COOLING COIL 107E5189/0 615 236 U41 C621C 3
17150 6.70 J.5 MCR HVAC SUPF FAN C 107E5189/0 615 237 U41 C621G 3
17150 6.70 J.6 MCR HVAC SUPP FAN G 107E5189/0 615 238 U41 DP1101C 3 17150 6.80 J.3 DIFF PRESS INDICATOR 107E5189/0 615 239 U41-F007C 3
17150 6.80 J.1 MO VALVE 107E5189/0 615 (pf 240 U41-F007
) 2.
17150 6.90 J1 MO VALVE 107E5189/0 615 241 U41-F008C 3
17150 6.80 J.1 MOVALVE 107E5189/0 615 242 U41-F011C 3
17150 6.60 J.1 MOVALVE 107E5189/0 615 U41- 0)8 J' >
17150 6.90 J.1 MO VALVE 107E5189/0 615 7
243 GPb 244 U41-ME104C 3 17150 6.60 J.1 MOISTURE ELEMENT 107E5189/0 615 245 U41 TE103C 3
17150 6.60 J.1 TEMP ELEMENT 107E5189/0 615 246 U41-DPl111 A 1 17150 5.20 K.8 DIFF PRESS INDICATOR 107E5189/0 619 247 U41-F104A 1
17150 5.20 K.8 MO VALVE 107E5189/0 619 248 U41 TE110A 1
17150 5.20 L.1 TEMP ELEMENT 107E5189/0 619 249 U41 F010F 2
17150 2.70 K.7 MO VALVE 107E5189/0 620 250 U41-POT 105F 2 17150 2.70 K.7 POSITION TRANSMITTER 107E5189/0 620 251 U41 C601B 2
17150 1.30 J.5 MCR HVAC SUPP FAN P 107E5189/0 621 252 U41-C601F 2
17150 1.30 J.6 MCR HVAC SUPP FAN F 107E5189/0 621 253 P25-F0058 2
17150 1.10 J.2 TCV: MCR CLG 107E5182/0 621 254 U41-B601B 2
17150 1.10 J.3 MCR COOUNG COIL 107E5189/0 621 255 U41-8601D 2
17150 1.10 J.3 MCR COOLING COIL 107E5189/0 621 256 U41 B601F 2
17150 1.10 J.3 MCR COOLING COIL 107E5189/0 621 257 U41-C603B 2
17150 1.80 J.1 MCR RECIRC SUPP FAN B 107E5189/0 621 258 U41 C603F 2
17150 1.80 J.1 MCR RECIRC SUPP FAN F 107E5189/0 621 259 U41 DPl101B 2 17150 1.10 J.3 DIFF PRESS INDICATOR 107E5189/0 621 260 U41-DPl106B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 261 U41-DP1107B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 262 U41 DPl108B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 263 U41-DP1109B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 2
17150 1.10 J.1 MO VALVE 107E5189/0 621 264 U41-F0078 4 d[o 265 U41-F007F 2
17150 1.10 J.1 MO VALVE 107E5189/0 621 266 U41-F008B 2
17150 1.10 J.1 MO VALVE 107E5189/0 621 g
C flo 267 U41.F008r 23 17150 1.10 J.1 MO VALVE 107E5189/0 621 9A.6 103 Fire Hazard Analysis Database
i l
Rev.0 ABWR Duign ControlDocumntmer2 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base-Sorted by Room-Control Building (Continued)
LOCATION LOCATION ITEM MPL ELECT ELEV.
NUMBER ALPHA SYSTEM ROOM NO.
NO.
DIV.
LOCATION COORD.
COORD.
DESCRIPTION DRAWING NO.
268 U41-F009B 2
17150 1.80 J.1 MO VALVE 107E5189/0 621 g
C Plo 269 U41-F009F
,E 3 17150 1.80 J.1 MO VALVE 107E5189/0 621 270 U41-F011 B 2
17150 1.50 J.1 MOVALVE 107E5189/0 621 271 U41-ME104B 2 17150 1.50 J.1 MOISTURE ELEMENT 107E5189/0 621 272 U41 TE1038 2
17150 1.50 J.1 TEMP ELEMENT 107E5189/0 621 273 P25-C0018 2
17150 2.80 J.4 dECW PUMP B 107E5182/0 623 274 P25-C001E 2
17150 2.80 J.8 HECW PUMP E 107E5182/0 623 275 P25-0001B 2
17150 2.80 J.4 HECW REFRIGERATOR B 107E5182/0 623 276 P25-0001E 2
17150 2.80 J.8 HECW REFRIGERATOR E 107E518U0 623 277 P25-DFT007B 2 17150 2.30 J.2 DP XMTR (FLO CONT B/E) 107E5182/0 623 278 P25-F0128 2
17150 2.50 J.2 PCV: HECW UNITS B/E 107E5182/0 623 279 P25-FIS003B 2
17150 2.80 J.4 FLOW IND SWITCH B 107E5182/0 623 280 P25-FIS003E 2
17150 2.80 J.8 FLOW IND SWITCH E 107E5182/0 623 281 P25-TE005B 2
17150 2.80 J.4 TEMP ELEM (UNIT B/E) 107E5182/0 623 282 P25-F016B,
2 17150 1.70 K.8 TCV: C/B ELEC RM B 107E5182/0 624 283 U41 B604B 2
17150 1.70 K.8 ESS EQUIP RM COOL COIL 107E5189/0 624 284 U41-8604F 2
17150 1.70 K.8 ESS EQUIP RM COOL Coll 107E5189/0 624 285 U41-C606B 2
17150 1.80 K.5 EM ELEC (B) SUPP FAN B 107E5189/0 624 286 U41 C606F 2
17150 1.80 K.6 EM ELEC (B) SUPP FAN F 107E5189/0 624
(~fb 287 U41 DP1111B 2 17150 1.60 K.8 D6F PRESS INDICATOR 107E5189/0 624j 288 U41 F104B 2
17150 1.60 K.8 MOVALVE 107E5189/0 624
[ 6 289 U41-TE110B 2
17150 1.60 L.1 TEMP ELEMENT 107E5189/0 624 290 U41 TE112B 2
17150 2.00 K.6 TEMP ELEMENT 107E5189/0 624 291 U41-C607B 2
17150 2.20 K.5 EM ELEC (B) EXH FAN B 107E5189/0 625 292 U41-C607F 2
17150 2.20 K.6 EM ELEC (B) EXH FAN F 107E5189/0 625 l
293 U41-F105B 2
17150 2.20 K.5 MO VALVE 107E5189/0 625 294 U41-TE1138 2
17150 2.20 K.5 TEMP ELEMENT 107E5189/0 625
)
295 U41-C602B 2
17150 2.70 K.5 MCR HVAC EXH FAN B 107E5189/0 626 296 U41 C602c 2
17150 2.70 K.*
MCR HVAC EXH FAN F 107E5189/0 626 297 U41-F010B 2
17150 2.70 K.5 MO VALVE 107E5189/0 626 A d A -> A97a U 4 8-F0 0(,
3 11150
.27o of haO VAW6 sp7t Stirf/o 62(a g4 298 U41.F012 B 2
17150 2.70 K.5 MO VALVE 107E5189/0 626 299 U41-POT 105B 2 17150 2.70 K.5 POSITION TRANSMITTER 107E5189/0 626 300 U41-C609C 3
17150 6.20 K.5 EM ELEC (C) EXH FAN C 107E5189/0 631 301 U41 C609G 3
17150 6.20 K.6 EM ELEC (C) EXH FAN G 107E5189/0 631 9A 6-104 FUe Hazard Analysis Database
i a
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e FIGURE 9.2-7 REACTOR SERVICE WATER SYSTEM P&lD (Sheet 1 of 3)
M*
ABWR DCD/ Tier 2 a v. o
S/RVs 3.4.2 SURVEILLANCE REQUIREMENTS
(
SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify the safety function lift setpoints
'" - -t 4 G of the required S/RVs are as follows:
k o.ccer [Ece W5 t
Number of Setpoint S/RVs (MPaG)
"I t, ; ~. e -_ Te~ sfing 2
7.92 0.0792
- f Prog r~
4 7.99 0.0799 's'
- i'iM*i' i
4 8.19 0.0819
.g cha y 6/90 Following testing, lift settings shall be
+.
within ! 1%.
c P to i
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam dome pressure is i
2 6.55 MPaG.
Verify each required S/RV opens when 18 months o a manually actuated.
Sf ea t alv solen l
)
ABWR TS 3.4-3 Rev. o. Design control oo vment/ Tier 2
i S/RVs B 3.4.2 l
BASES ACTIONS The 14 day Completion Time to restore the inoperable (continued) required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event ri.iing S/RV actuation, and a reasonable time to complete the Required Action.
B.1 and B.2 l
With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the i
ASME Code limit on reactor pressure.
If the inoperable required S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1 or if two or more required S/RVs are inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ar.d to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
l SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This Surveillance demonstrates that the required S/RVs will open at the pressures assumed in the safety analysis of
~
Reference 2.
The demonstration of the S/RV safety function lift settings must be performed during shutdown, since this is a bench test. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
The S/RV setpoint is 3% for OPERABILITY; however, the valves are reset to 1% during the Surveillance to allow for drift.
2: IS = nth frequency w : : ic ted b:::::: thi:
%-veilloma must. be crf:med dur%g :hutde,wn conditia :
rd i: 5:::d :n the time betu::: e%eliar e a cc o rh* C"--
Sfteifical fQ fyg aeue ib
- 1
~,y u r-,e, a 1p n p y Tc 5 Fi CIIO 6l6l4b W i
i r (continued)
L ABWR TS B 3.4-6 new. c. o sign centrol ooc - t m er z