ML20070Q525
| ML20070Q525 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/1989 |
| From: | Lobner P, Wooten B SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| References | |
| CON-FIN-D-1763, CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1011, NUDOCS 9103290110 | |
| Download: ML20070Q525 (115) | |
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NUCLEAR POWER PLANT g
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O BRUNSWICK 1 & 2 50 325 and 50 324 o
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SYSTEM SOURCEBOOK
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n BRUNSWICK 1 & 2 50 325 and 50 324 Editor: Peter Lohner Author: Ilruce Wooten
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Prepared for:
U.S. Nuclear Itegulatory Commission Washington, D.C.
20555 Contract NRC 03 87 029 FIN D 1763 Ly2WC
Brunswick 1 & 2 I
- rAllLE OF CONTENTS s.
au 1
S UhthtA R Y D ATA ON PLA NT............................................
1 2
IDENTIFICATION OF Siht!LAR NUCLEAR POWER PLANTS....
1 3
S YSTE h1 INFO RhiATIO N.............................................*.....2 3.1 Reactor Coolant S yste m (RCS)................................
8 3.2 Reactor Core Isolation Cooling (RCIC) System.............
13 3.3 Emergency Core Cooling System (ECCS)...................
18 3.4 Instrumentation and Control (I&C) Systems.................
33 3.5 Elee tric Powe r S yste m.........................................
37 3.6 Control Rod Drive liydraulle System (CRDliS)............
59 3.7 Service Water System (SWS) and Residual licat Removal hervice Water System (RSWS)...............
62 4
PL A NT I N FORh 1 ATION....................................................
68 4.1 Si te and B uildin g S ummary....................................
6S 4.2 Facility Layou t Drawing s.......................................
68 5
BIBLIOG R APIIY FOR BRUNSWICK 1 AND 2........................
96 O
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APPENDIX A, Definition of Symbols Used in the System and Layou t Dra win g s..........................................
97 APPENDIX B. Definition of Terms Used in the Data Tables...........
104 i
1/89 i
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Brunswick 1 & 2 p
LIST OF FIGURES figme Esuu:
31 Cooling Water Systems Functional Diagram for Brunswick 1 and 2....
7 3.1-1 B runswick 2 Reactor Coolant System..................................... 10 3.1 2 Brunswick 2 Reactor Coolant System Showing Component loc a ti o n s.............................................. <.......................... Il s
3.2 1 Brunswick 2 Reactor Core Isolation Cooling System.....................
15 1
3.2 2 Brunswick 2 Reactor Core Isolation Cooling System Showing Component Loeattons..........................................................
I6 3.3 1 Brunswick 2 liigh Pressure Coolant injection System................... 22 Brunswick 2 liigh Pressure Coolant In 3.3 2 Showing Component Locations.........jection System 23 3.3-3 B runswick 2 Core S pray Syst em............................................ 24 3.3 4 Brunswick 2 Core Spray System Showing Component Loeatons..........................................................
25 V
3.3-5 Brunswick 2 Residuallleat Removal System, Loops A and C..........
26-3.3-6 Brunswick 2 Residual 11 eat Removal S Showing Component Locations..........ystem. Loops A and C 27 3.3 7 Unmswick 2 Residual Heat Removnl System. Loops B and D..........
28 3.3 8 Brunswick 2 Residuallieat Removal S Showing Component Locations..........ystem, Loops B and D 29 l
3.5 1 Brunswick 1 & 2 4160 VAC Electric Power Distribution System......
40 3.5 2 Brunswick 1 & 2 4160 VAC Electric Power i
Distribution System Showing Component Locations......................
41 3.5 3 Brunswick 14160/480 VAC Electric Power System..................... 42 3.54 Bnmswick 14160/480 VAC Electric Power S Component Locations............................ystem Showing
............................43 3.5 5
. Brunswick 2 4160/480 VAC Electric Power System..................... - 44 3.5 6 Brunswick 2 4160/480 VAC Electric Power S Compone nt Locations.............................yste m S howing 45 O
Q 3.5 7 Simplified One Line Diagram of Brunswick 1 and 2125/250 VDC Elec tri c Power Syst e m......................................................... 46 4
il 1/89
-- ~. -
Brunswick 1 & 2 LIST Of FIGURES (continued)
J Fica East 3.5 8 Simplified One Line Diacram of Brunswick I and 2125/250 VDC Electric Power System Showing Component Locations.................. 47 3.5 9 Typical Three Line Diagram of Brunswick 1 and 2125/250 VDC Electric Power S ys t e m........................................................ 48 3.5 10 Brunswick 1 and 2 120 VAC Electric Power System......................
49 3.5. I 1 Brunswick 1 and 2120 VAC Electric Power System Showing 3
Component Loeations..........................................................
50 j
3.6 1 Simplified Diagram of Portions of the Control Rod Drive fl System That Are Related to the Scram Function...........ydraulle 61 3.7 1 Brunswick 2 Service Water System 65 3.7-2 Brunswick 2 Service Water System Showing Component Locations... 66 4-1 General View of the Brunswick Site and Vicinity.........................
70 4-2 Brunswick I and 2 Simplified Site Plan...................................
71 b
43 Brunswick Reactor Building Section, Looking West....................
72 4-4 Brunswick 1 Reactor Building. Control Building and Radwaste B uildin g, Ele vation 17 '0"...................................................
73 45 Brunswick 1 Reactor Building. Elevation 20'0" and Control Building and Radwaste Building, Elevation 23'0"........................
74 Brunswick 1 Reactor Building, Elevation 50'0" and Control Buildin 4-6 and Radwaste Building, Elevation 49'0"................................. g 75' 47 Brunswick 1 Reactor Building. Elevation 80'0" and Control B u ild in g. Ele va ti on 7 0'0".................................................... 76 48 Brunswick 1 Reactor Building, Control Building, Elevation 117'4"..........................................................................
77 49 Brunswick 2 Reacter Building, Elevation 17'0".........................
78 4-10 Brunswick 2 Reactor Building. Elevation 20'0".........................
79 4 11 Brunswick 2 Reactor Building, Elevation 50'0"..........................
80-4 12 Brunswick 2 Reactor Building, Elevation 80'0"..........................
81 4 13 Brtmswick 2 Reactor Building, Elevation 117'4"........................
82 lii 1/89 i
... ~ - _.
Brunswick 1 & 2 LIST OF FIGURES (continued)
Fipre M
4 14 Brunswick 1 and 2 Diesel Generator Building, Elevation l'6"........
83 4 15 Brunswick 1 and 2 Diesel Generator Building, Elevation 23'0"........
84 4 16 Brunswick i and 2 Diesel Generator Building, Elevat!on 50'0"........
85 4 17 Brunswick 1 and 2 Service Water Intake Structure, Elevation 13'4"..
86 4 18 Brunswick I and 2 Service Water intake Structure, Elevation 20'0"...
87 A1 Key to Symbols in Fluid System Drawings................................
100 A-2 Key to Symbo's in Electrical System Drawings........................... 102 A3 Key to Symbois in Facility Layout Drawings..............................
103 i
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i Brunswick 1 & 2
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LIST OF TABLES Tahk Eagg 31 Summary of Brunswick 1 and 2 Systems Covered in this Repon.......
3 3.1 1 Brunswick 2 Reactor Coolant System Data Summary for S elec ted Componen ts...................................................... 12 3.2 1 Bmnswick 2 Reactor Core Isolation System Data Summary for S elec t ed Co mpone n t s...................................................... 17 l
3.3 1 Brunswick 2 Emergency Core Cooling System Data Summary for S e lec t ed Compon en t s...................................................... 30 3.5 1 Brunswick 2 Electric Power System Data Summary for S e lect ed Compon en ts...................................................... 51 3.5 2 Partial Listing of Electrical Sources and Loads at Brunswick2..........
53 3.7 1 Brunswick I and 2 Service Water System Data Summary for S el ec t ed Compon e n t s...................................................... 67 41 Definition of Brunswick 2 Building and Location Codes................. 88 4-2 Partial Listing of Components by Location at Brunswick 2,..............
90 B-1 Compon e n t Type Code s...................................................... 105
(
4 v
-1/89
Brunswick 1 & 2 CAlmnN The information in this repon has been developed over an extended period of time based on a site visit, the Final Safety Analysis Report, system and layout drawings, and other published infonnation. To the best of our knowledge, it accurately reflects the plant configuration at the time the infonnation was obtained, however, the infonnation in this document has not been independently verified by the licensee or the NRC.
NOT1CE
'Diis sourcebook will be periodically u 3 dated with new and/or replacement pages as appropriate to incorporate adcitionalinfonnation on this reactor plant, Technical errors in this report should be brought to the attention of the following:
hit. hiark Rubin U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation -
Division of Engineering and Systems Technology hiail stop 7E4 Washington, D.C. 20555 With copy to:
hir. Peter Lobn:r hianager, Systems Engineering Division Science ApplicationsInternationalCorporation 10210 Campus Point Drive San Diego, CA 92131 (619)458 2673 Correction and other recommended changes should be submitted in the fonn of marked up copies of the affected text, tables or figures. Supporting documentation should be included if possible.
Og "vi 1/89-
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1 4
IlllUNSWICK 1 & 2 l
ItECOllD OF ItEVISIONS 11EVISION ISSUE COhlh1ENTS 0
1/89 Original report i
9 i
4 1
i l
vii 1/89
Brunswick 1 & 2 l
IIRUNSWICK 1 AND 2 SYSTEh! SOURCEllOOK This sourcebook contains summary information on Drunswick 1 and 2. Summary data on this plant are presented in Section 1. and similar nuclear power plants are identified in Section 2. Information on selected reactor plant systems is presented in Section 3, and the site and building layout is illustrated in Section 4. A bibliography of reports that describe features of this plant or site is presented in Section 5. Symbols used in the system and layout drawings are defined in Appendix A. Terms used in dat. tables are defined in Appendix B.
4 1.
SUNIN1ARY DATA ON PLANT Basic information on the lirunswick nuclear plant is listed below:
Docket number 50 325 (Unit 1) and 50-324 (Unit 2)
Operator Carolina Power and Light Company location Brunswick County in North Carolina Commercial operation date 3n7 (Unit 1),1In$ (Unit 2)
Reactor type BWR/4 NSSS vendor General Electric Power (hlWt/h1We) 2436/821 Architect engineer United Engineers and Constructors Containment type Steel drywell and wetwell (hlark 1) 2.
IDENTIFICATION OF Sih!!LAR NUCLEAR POWER PLANTS The Brunswick nuclear plant has two General Electric BWR/4 nuclear steam supply systems on the site. Each unit has a Mark I BWR containment incorporating the drywell/ pressure suppression concept, and has a secondary containment structure of reinforced concrete, Other BWR/4 plants in the United States are as follows:
Vemiont Yankee Drowns Ferry Units 1,2 and 3 llatch Units 1 and 2 Cooper Nuclear Station Fitzpatrick Duane Arnold Peach Bottom 2 and 3 Fermi Unit 2 Hope Creek Unit 1 Limerick Units 1 and 2 (hlark Il Containment)
Shoreham (hlark 11 Containment)
Susquehanna Units 1 and 2 (hlark U Containment)
The Brunswick plants have a high pressure coolant injection system, a reactor core isolation cooling system capable of steam condensing operation in conjunction with the RHR system, a low pressure core spray system and a multi mode RHR system.
l 1
1/89 4
1
Brunswick 1 & 2 1
j 3.
SYSTEM INFORMATION This section contains descriptions of selected systems at Brunswick I and 2 in terms of general function, operation, systern success criteria, major components, and L
support system requirements. A summary of major systems at Brunswick 1 and 2 is presented in Table 31. In the " Report Section" column of this table, a section reference (i.e. 3.1,3.2, etc.) is provided for all systems that are described in this report. An entry of "X" in this column means that the system is not described in this report. In the "FSAR 4
l Section Reference" column, a cross reference is provided to the section of the Final Safety Analysis Report where additional information on each system can be found, Other sources ofinformation on this plant are identified in the bibliography in Section S.
Several cooling water systems are identified in Table 31. The functional relationships that exist among coolinj: water systems required for safe shutdown are shown j
in Figure 31. Details on the indivic ual cooling water systems are provided in the report sections identified in Table 31.
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Summary of Itrumswick I amt 2 Systems Cmcrcel in this Regmrt Generie Plant-Specific Report FSAR Secti<m System Name Ststem Name Sectiem Reference Reactor IIcat Removal Systems Reactor Coolant System (RCS)
Same 3.1 5
Reactor Core Isolation Cooling Same 3.2 5.4.6 (RCIC) Systems
-. Emergency Core Cooling Sys* ems Same (ECCS)
Iligh-Pressure Injection Ifigh-Pressure Coolant injection 3.3 6.3
& Recirculation (IIPCI) System
- Low-pressure Injection Core Spray (CS) System, 3.3 6.3
& Recirculation Iow-Iwssure Coolant Injection 3.3 6.3 (LPCI) System (an operating med of the RIIR system) l w
1
- Automatie Depressurization Same 3.3 6.3 System (ADS)
- Decay IIcat Removal (DIIR)
Residual Heat "emoval 3.3 5.4.7. 6.3 System (Residual IIcat Removal (RIIR) System (a multi-mode (RIIR) System) system)
Main Steam and Power Conversam Main Steam Supply System, X
10.2, 10.3, 10.4.4 Systems Condensate System, X
10.4.7 Feedwater System, X
I0.4.7 Condenser CirculatingWater X
10.4.5 System
- O:herIIcat Removal Systems Steam-condensing RIIR/RCIC 3.2 5.4.6, 5.4.7 operation C-E
O.
-x w
w Table 3-1.
Summary of Ilrtmswick I aml 2 Systems Covered in this Report (Continued)
Generic Plant-Specific Report FSAR Section l
System Name Sutem Name Section Reference t
Reactor Coolant Inventory Control Systems Reactor Water Cleanup (RWCU)
Same X
5.4.S l
System ECCS See ECCS,above Control Rod Drive Ilydraulic System (CRDIIS) Same 3.6 4.6 i
Containment Systems
- Primary Containment Same(dryweII and pressure X
6.2 suppression chamber)
- Secondary Containment Same X
6.2 A
- Standby Gas Treatment System (SGTS)
Same X
6.5.1.1
- Containment IIcat Removal Systems
- Suppression Pool Cooling System Same (an operating mode of the 3.3 6.2.2 RIIR system)
Containment Spray System Same(anoperatingmodeof the 3.3 6.5.2 i
RIIR system)
- Containment Fan Cooler System Sane X
6.2.2
- Containment Normal Ventilation Systems Containment Fan Cooler System X
6.2.2 Combustible Gas Control Systems Containment Atmospheric Control X
6.2.5 System, Containment Attmspheric Dilution System Containnent Inening System,Codnmem Atmospheric i
Make-up System, Liquid Nitrogen e-Supply to Augmemed Off-gas 3
Charcoal AbsorberSystem t
'l Tahic 3-1.
Summary of Hitmswick I and 2 Systems Cavered in this Regmrt (C mfinned) l
)
R Generic Plant-Specific Report FSAR Sectiem System Name S5 stem Name Section Reference i
Reactor and Reactivity Control Systems Reactor Core '
Same X
4 l
- Control Rai System Control Rod Drive System X
a
- Chemical Poison System S: r.dby Liquid Control System X
9.3.4 (SLCS)
Instrumentation & Control (I&C) Systems Reactor Protection System (RPS)
Same 3.4 7.2 i
- Engineered Safety Itature Actuation Engineered Safety Feature Systems X-7.3 i
System (ESi'AS),
(including primary containment i
isolation and nuclear steam supply i
u shutoff system, secondary
[
containment isolation and standby j
gas treatment system actuation, and core standby cooling system -
actuation)
- Remote Shutdown System Ieca! control panels
.3.4 7.4 p
OtherI&C Systems Various othersystems.
X 7.5 to 7.7 Support Systems -
j Class 1E Electric IbwerSystem Same 3.5 8.2,83 j
i
- Non-Class 1E Electric Power System Same X
3.2,83 j
Diesel Generator Auxiliary Systems.
Same.
3.5
- 83. 9.5.4.9.5.6 i
o 9.5.7. 9.5.8 i
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Summary of Brunswick I and 2 Systems Covered in this Report (Continued)
Generic Plant-Specific Report FSAR Section System Name S5 stem Name Lettion Reference Support Systems (continued)
- Component Couling Water (CCW)
Reactor Buik!ing Clmer!
3.7 9 2.2 System Conling Water (RBCCW) Syst m
- Service Water System (SWS)
Service Waterand RIIR 3.8 9.2.1. 9.2.5 Service Water ILester System
- ResidualIIcat Removal Senice Water Service Waterand RIIR 3.8 9.2.1. 9.2.5 (RIIRSW) System Service WaterIkmterSystem i
- Other Cooling Water Systems Turbine Buikling Cl<rei X
9.2.7 Cooling Water frBCCW) System i
es Make-up WaterTreatment System X
9.2.3 Fire Protection Systems Same X
9.5.1 i
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- Room Heating. Ventilating,and Air-IIabitability System.
X 6.4 Conditioning (IIVAC) Systems Air-Conditioning. IIcating.
X 9.4 Cooling and Ventilation Systems j
- Instrument and Service AirSystems Compcned Air System X
9.3.1 f
Refueling and Fuel Storage Systems Same X
9.1 Radioactive Waste Systems Same X
11 Radiation Protection Systems Same X
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Brunswick 1 & 2 3.1 REACTOR COOLANT SYSTEM (RCS) 3.1.1 Svetem Funclinn The RCS. also called the Nuclear Steam Supply System (NSSS),is responsible for directing the steam stoduced in the reactor to the turbine where it is used to rotate a generator and produce e ectricity. The RCS pressure boundary also establishes a boundary against the uncontrolled release of radioactive material from the reactor core and primary coolant.
3.1.2 System tiefinitio11 The RCS includes: (a) the reactor vessel, (b) two recirculation loops, (c) two recirculation pumps (d) 11 safety / relief valves, and (c) connected piping out to a suitable isolation valve boundarv. Simplified diagrams of the RCS and important system interfaces are shown in Figt.res 3.'l-1 and 3.12. A summary of data on selected RCS components is presenNd in Table 3.1-1.
3.1.3 Svstem 'Joerntion During p)wer operation, circulation in the RCS is maintained by one recirculation pump it each of the two recirculation loops and the associated jet pumps internal to the reactor vessel. The steam water mixture flows upward in the core to the steam dryers and sepa ators where the entrained liquid is removed. The steam is piped through the main steam lines to the turbine. The separated liquid returns to the core, mixes with the feedwater and is recycled again.
A ponion of the liquid in the downcomer region of the reactor vessel is drawn off by the recirculation pumps. The discharge of these pumps is rettwned to the inlet nozzles of the jet pumps at high velocity. As the liquid enters the jetrumps, the slow O
moving liquid m tie upper region of the downcomer is induced to flow through the jet pumps, producing reactor coolant circulation.
The steam that is produced by the reactor is piped to the turbmc via the main steam lines. There are two main steam isolation valves (MSIVs) in each main steam line.
Condensate from the turbine is returned to the RCS as feedwater, l
Following a transient that involves the loss of the main condenser or loss of feedwater, heat from the RCS is dumped to the suppression chamber via safety / relief valves on the main steam lines. A LOCA inside containment or operation of the Automatic Depressurization System (ADS) also dumps heat to the suppression chamber. Makeup to the RCS is provided by the Reactor Core Isolation Cooling (RCIC) system (see Section 3.2) or by the Emergency Core Cooling System (ECCS, see Section 3.3). Heat is transferred from the containment by the Residual Heat Removal (RHR) System operating in the containment cooling mode. The Service Water System completes the heat transfer path from the containment to the ultimate heat sink (see Section 3,7) Actuation systems 3rovide for automatic closure of the MSlVs and isolation of other lines connected to the ACS.
RCS overpressure protection is provided by eleven safety / relief valves which discharge to the suppression pool.
3.1.4 System Success Crlferia The RCS success criteria can be described in terms of LOCA and transient mitigation, as follows:
An unmi:igatible LOCA is not initiated.
If a mitigatible LOCA is initiated, then LOCA mitigating systems are successful.
p I
8 1/89-40
Brunswick 1 & 2 Ji a transient is initiated, then either:
RCS integrity is maintained and transient mitigating systems are successful, or RCS integrity is not maintained, leading to a LOCA like condition (i.e.
stuck open safety or relief valve, reactor coolant pump seal failure), and LOCA mitigating systems are successful.
3.1.5 Comnonent Informntl2D A. RCS
- 1. Steam flow: 10.47 x 106 lb/hr
- 1. Nonnal operating pressure: - 1005 psig B. Safety / relief Valves (11)
- 1. Set pressure: 4 @ l105 psig,830.000 lb/hr (each) 4 @ l115 psig. 838,0 lb/hr (each) 3 @ l125 psig 845,000!b/hr(each)
C. Recirculation Pumps (2)
- 1. Rated flow: 45.200 rpm @ unknown head
- 2. Type: Vertical centnfugal D. Jet Pumps (20) 3.1.6 Suonort Systems and Interfaces A. hiotive Power l. The recirculation pumps are supp!!cd by non Class lE power.
B. hlSIV Operating Power The instrument air system supports nonnal operation of the hiSIVs. Valve operation is controlled by an AC and a DC solenoid pilot valve Both solenoid valves must be deenergized to cause blSIV closure. This design prevents spurious closure of an hISIV if a single solenoid valve should fall. htSIVs are designed to fall closed if the pneumatic supply is lost or if both AC and DC control power is lost to the solei, aid pilot valves, This is achieved by a local-dedicated air accumulator for eac'1 htSIV and an independent valve closing spnng.
C. Recirculation Pump Cooling The Service Water System provides cooling water to the recirculation pump coolers.
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Brunswick ? Reactor Coolant System Showing Component Locations t
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Brunswick 2 Reactor Coolant System Data Summary for Selected Components l
1 J
COMPONEF4T !D COMP.
LOCAT!Off POWER SOURC'E VOLTAGE POWER SOURCE E 18 E R G.
' TYPE LOCATIOff LOAD GRP.
HPCI-2 MOV HC EP-MCC-2XD 460 20HB AC!E4 HPCI-3 MOV ECCSitJL EP-MCC-2XDA 125 20HB OC/2A lHCIC-7 MOV HC EP-MCC-2XC 480 20HB AC/E3 HGC B MOV ECCSirJL EP-MCC-2XDB 125 20HB OC/28 HCS-1 MOV HC EP-MCC-2XC 480 2UHB AC/E3 I
HCS-16
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HHH-8 MOV ECCSIFJL EP-MCC-2XDU j 125 20HB DC/28 l HHH-9 MOV HC
/-MCC 2XA 480 20HB AC,E3 t
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Brunswick 1 & 2 3.2 REACTOR CORC ISOLATION COOLING (RCiC) SYSTEM 3.2.1 Sntem Function The maetor core isolation cooling system provides adequate ec,rt cooling in the event that reacu im ation 5 accompanied by loss of feedwater flow. This system provides makeup at reactoru 3erating pressure and does not require RCS deornscrization.
The RCIC system is nca ci nsidered to be part of the Emergency Core Cooliig System (ECCS. See section 3.3) au does tot have a LOCA mitigating function.
3.2.2 Svetem DefinitlM The reactor core is tation cooling syt'ern consists of a steam turbine driven pump and associated valves ami piping for delivenng makeup water from the condensate story:e tank or the suppressi, poe! a the reactor pressure vessel. The RCIC can also operate h motion with ; Rilk system in the steam condensing mode,in which condensed del!.ered iom the RilR heat exchanger outlets to the RCS pump suction, fu u ac s the RCS.
Tne RClO turbine is driven by steam from main steam line C. The tche exhausts to the suppression pool, Simplified drawings of the reactor core isolation cooling system are show %
Ficures 3.21 and 3.2 2. Interfaces between the RCIC and the RCS are shown in Sectico 3.I. A summary of data on.,cyted RCIC system components is presented in Table 3.21.
i 3.2.3 Snteln On Silull Deri, g normW Wratioi, 4 a RCIC is in standby with tf steam supply valvet so the RCIC turbine driven puap cle 5 i and the pump suction abgned to the conde.isate storage tank.
3 Upon rc@t of an RPV low water level signal, the turbine sump steam supply i
salves are opened anc makeup wuer is supplied to the RPV via fecc wr.ter line B. The prirtary wa:er supply for the RCIC is the conelensate storage tank (CST). The suppression pool is used as a backum water supply Reactor core heat is dumped to the suppression pool via the safety / relief valves which cycle as needed to limit RCS pressure. The RCIC turbine also exhausts to the suppression pool.
The RCIC can aise operate m conjunction with the RIIR system in the steam condensing mode, b @ht condensed steam is delivered from the RHR heat exchanger outlets to the FGC pump etian, for return to 'the RCS in this mode of operation, reactor core but is transfemd,c. +e R HR system eather than to the suppression pool. The i
RCIC turbirm still exhausts to the s 'tyjression pc '.
l 3.2,4 System Succeu Criterig bor the RCIC system to be q%.'sful there must be at least one water source I
and supply path to the turbine. driven purgh..icpen steam supply path to the turbine, an open discharge path to the RCS, and an open tunine exhaust path to the suppression pool.
3.2.5 Demnnnmt informatinn A. Steam turbine driven RCICpump:
- 1. Rated Flow: 400 gpm @ 2,800 f head (1,214 psid)
- 2. Rated Capacity: J00%
- 3. Type: centrifugal B. CoaJensate Storage Tank
- 1. Capacity: 200.000 gal (reserved for RCIC use)
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A. Control Signals
- 1. Automatic
- a. The RCIC pump is automatically actuated on a reactor vessellow water level signal. The system automatically shuts down when the reactor vessel water level reaches a specified Ievel, and automatically restarts if the level retums to the low level trip point.-
- b. The RCIC pump suction is automatically switched to the suppression pool upon low condensate storage tank level,
- c. The RCIC pump is automatically shutdown upon receipt of any of the following signals:
Reactor vessel high water level Turbine over speed Pump low suction pressure Turbine high exhaust pressure
- 2. Remote Manual The RCIC pump can be actuated by remote manual means from the control room or the remote shutdown panel, Manual action is required to place the RCIC in the RHR steam condensing mode.
B. Motive Power
- 1. The RCIC turbine driven pump is supplied with steam from main steam line C, upstream of the main steam isolation valves.
T
- 2. The RCIC motor operated valves are either Class IE AC or Class 1E DC loads that can be supplied from the standby diesel generators or the sta: ion batteries, respectively, as described in Section 3.5.
C. Other
- 1. Lubrication and cooling for the turbine driven pump are assumed to be supplied locally.
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A Table 3.2-1. __ Brunswick 2 Reactor Core isoloation Cooling System Da:a Summary fOr Selected Components COMPOtJENT ID COMP.
- LOCATIOt2 POWER SOURCE VO LT A GE POWER SOURCE EMERG.
TYPE LOC A TION LOAD GRP.
CST TAtJK CSI RCIC IDP RHRBRM flCIC-10 MOV Rfi'.BRM EP-MCC-2XDB 125 20RB DC/28 f1CIC-11 MOV...
RHRBRM EP-MCC-2XDB 1 5
'>0RB JC/28 flCIC-12 MOV RHRBRM EP-MCC-2XDB 125 i20RB DG/2B flC!C-13 MOV 20HB EP-MCC-2XDB 125 20RB DC/28 flCIC-19 MOV RflHBRM.
EP-MCC-2XDB 125 20RB DC/2B
~
flCIC-22 MOV flHRBRM EP-MCC-2XDB 125 20RB DC/28 RCIC-29 MOV RHRBRM EP-MCC-2XDB 125 20RB DC/2B RCIC-31 :--
MOV RHRBRM EP-MCC-2XDB 125 20RB DC/2B RCIC-45 MOV-RHRBRM EP-MCC-2XDB 125 20RB DC/28 RCIC-46 MOV flHRBRM =
EP-MCC 2XDB 125 20RB DC/2B RCIC-7 MOV' RC:
EP-Mt.,G-2XC 480 20HB AC/E3 RCIC MOV ECCSINL '.
EP-MCC-2XDB 125 20RB DC/28 RCIC-TTV8.
MOV RHRBRM EP-MCC-2XDB 12f 20RB DC/2B RCS-328 MOV; MSIVRM EP-BS-E8 480 23DGSG8 AC/E4 r
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Brunswick 1 & 2 3.3 O
EMERGENCY CORE COOLING SYSTEM (ECCS) 3.3.1 Sutem Function The ECCS is an integrated set of subsystems that perform emergency coolant injection and recirculation functions to maintain reactor core ccolant inventory and adequate decay heat removal following a LOCA. The ECCS also performs suppression pool cooling and containment spray functions and has a capability for mitigating transients.
3,3.2 Svetem Definition The emergency coolant injection (ECI) function is performed by the following ECCS subsystems:
High Pressure Coolant Injection (HPCI) System Automatic Depressurization System (ADS)
Core Spray (CS) System Low Pressure Coolant Injection (LPCI) Syste.n The HPCI system is provided to supply make up water to the reactor pressure -
vessel (RPV) in the event of a small break LOCA which does not result in a rapid depressurization of the reactor vessel. The HPCI system consists of a turbine driven pump, system piping, valves and controls. The HPCI pump can draw suction from either the CST or the suppression pool. Water is injected into the reactor via feedwater line A.
The HPCI turbine is driven by steam from main steam line A. The turbine exhaust:; to the suppression pool.
The automatic depressurization system (ADS) provides automatic RPV O
depressurization for small breaks and transients so that the low pressure systems (LPCI
. Q and CS) can provide makeup to the RCS The ADS utilizes 7 of the 11 safety / relief valves that discharge the high pressure steam to the suppression pool.
The CS system supplies make up water to the reactor vessel at low pressure.
The system consists of two motor driven pumps to supply water from'the suppression pool to two spray spargers in the reactor vessel above the core.
The low pressure coolant injenion system is an operating mode of the residual heat removal (RHR) system, and provides make up water to the reactor vessel at low pressure. The LPCI system consists of two loops with two pumps in each loop. The four pumps with their associated loops are designated LPCIA, LPCIB, LPCIC, and LPCID.
Each loop consists of a motor driven pump which supplies water from the suppression poolinto the reactor vessel. There are two heat exchangers in the system, one for pumps 2A and 2C and one for pumps 2B and 2D The RHR system can be manually realigned as needed to perform suppression pool cooling or containment spray as part of the basic emergency core coohng function. The RHR system also can be ahgned for steam condensing operation, where steam from the HPCI steam line is condensed in the RHR heat exchangers, then piped to the suction of the RCIC pumps for the return to the reactor.
This is not an ECCS function.
Simplified drawings of the HPCI system are shown in Figures 3.3-1 and 3.3-
- 2. The CS system is shown in Figures 3.3-3 and 3.3 4. The LPCI system is shown in Figures 3.3 5 and 3.3 6 (loops A and C), and Figures 3.3 7 and 3.3 8.(loops B and D).
Interfaces between these systems and the RCS are shown in Section 3.1. A summary of selected data on ECCS components is presented in Table 3.31 3.3.3 Sntem Oneration
[.
All ECCS systems normally are in standby. The manner in which the ECCS operates to protect the reactor core is a function of the rate at which coolant is being lost
\\
from the RCS. The HPCI system is normally aligned to take a suction on the Condensate 18 1/89
i Brunswick 1 & 2 O
Storage Tank (CST). The HPCI system is automatically started in response to decreasing RPV water level, and svill serve as the primary source of makeup if RCS pressure remains high. Reactor core heat is dumped to the su which cycle as needed to limit RCS pressure.ppression pool via the safety / relief valves, Steam to drive the HPCI turbine is routed from main steam line A. If the break is of such a size that the coolant loss exceeds the HPCI system capacity, then the CS and LPCI systems can provide higher capacity makeup to the reactor vessel.
The Automatic Depressurization System will automatically reduce RCS pressure if a break has occurred and RPV water levelis not mamtained by the HPCI system. Rapid depressurization permits flow from the CS or LPCI systems to enter the vessel.
The CS system consists of two loops, each containing one 100% capacity pump. Each loop provides makeup to the reactor vessel through separate spray spargers.
The source of water is the suppression pool.
The LPCI system is an operating mode of the RHR system. In the LPCI mode the four pumps take suction on the suppression pool and inject back into the vessel through the reactor recirculation loops. Other operating modes of the RHR system include: (a) suppression pool cooling, in which water is recirculated from the suppression pool through two RHR heat exchangers and back to the suppression pool; (b) containment spray,in which water is pumped to fog jet nozzles in the drywell and suppression pool: (c) steam condensing, in which condensate is delivered to the RCIC pump; and (d) shutdown-cooling.
3.3.4 Sutem Success Criterin LOCA mitigation requires that both the emergency coolant injection (ECI) and emergency coolant recirculation (ECR) functions be accomplished. The success criteria are not clearly defined in the Brunswick 1 and 2 FSAR but can be inferred from pump h
capacities that are def'ined based on certain design basis accidents that are considered in the
'v/
licensing process. On this basis, the ECI system success criteria for a large LOCA are the following:
1 of 2 core spray pumps with a suction on the suppression 30o1, or 3 of the 4 low pressure coolant injection _ pumps wit 1 a suction on the suppression pool.
The ECI system success criteria for a small LOCA are the following:
The high pressure coolant injection (HPCI) pump with a suction on the suppression poel or the condensate storage tank, or The automatic depressurization system (ADS) and 3 of 4_ LPCI pumps w.h a suction on the suppression pool, or The automatic depressurization system and 1 of 2 core spray pumps with a suction on the suppression pool.
The success criterion for the ADS is the use of any 1 of 2 ADS trains Note that there may be integrated success criteria involving combinations of core spray and LPCI pumps, it is i
possible that the coolant inventory control function for some small LOCAs can be satisfied l
i by low capacity high-pressure injection systems such as the control rod drive hydraulic
{
system (see Section 3.6).
The ECR success critena for LOCAs are integra,ed with the ECI success criteria-above..All injection systems essentially are operating in a recirculation mode when A{jN drawing water from the suppression pool.
19 1/89
Brunswick 1 & 2 For transients, the success criteria for reactor coolant inventory control involve the following:
Either the reactor core isolation cooling (RCIC) system (not part of the ECCS, see Section 3.2), or Small LOCA. mitigating systems For the sup 3ression pool cooling function to be successful, one of two
~
RHR trains must be alignec for containment heat removal and the associated RHR service water train must be operating to complete the heat transfer path from the RHR heat exchangers to the ultimate heat sink. In a given RHR train, one of two pumps must i
operate.
3.3.5 Comnonent Informntion A. Motor driven HPCI pump Pl 1 Rated flow: 4250 gpm @ unknown head
- 2. Rated capacity: 100 %
3
- 3. Type: centrifugal B. Motor-driven CS pumps 2A,2B
- 1. Rated flow: 4700 gpm @ l13 psid (vessel to'drywell) -
t
- 2. Rated capacity: 100 %
1
- 3. Type: centrifugal l
C. Motor driven LPCI pumps 2A,2B,2C,2D 1, Rated flow: :7,700 gpm @ 20 psid (vessel to drywell) '
- 2. Rated capacity: 331/3%
3, Type: centrifugal =
D,4 RHR' Heat Exchangers.2A and 2B :
1, Heat transfer capability:-90.8 x 106 BTU /hr-
- 2. Rated capacity:- 100%
- 3. Type: shelland tube I
E. Automatic-depressurization valves (7) -
- 1. Rated flow: 825,000 lb/hr @ l125 psid -
- 2. Ratedcapacity: 16.7 %
F, Pressure Suppression Chamber
- 1. Design temperature:'220'F
- 2. Maximum operating temperature: 95 F(assumed)!
- 3. Minimum water volume: 87,600 ft3 3.3.6-Sunnnrt %temt nnd Interrnees-3 A. Control signals
- 1. Automatic
- a. The HPCI pump, CS' pumps, and-the LPCI pumps, and allztheir associated valves function upon receipt oflov/ water levelin the reactor
)
Z 20 1/89-
_ _. _ _ _. ~. _. _.. _..
Brunswick IL& 2-vessel or high pressure in the drywell; When the reactor vessel pressure d
is low enough, the CS and LPCIinjection valves open.
- t
j reactor vessel high water level, HPCI pmap low suction pressure, or.
HPCI turbine exhaust high ?ressure.-. If an initiation signal is received after the turbine is shut c own, the system restarts automatically,-
i provided no shutdown signal exists.
I
- c. HPCI pump suctiowis automatically switched from the CST to the-suppression pool upon low CST level or high suppression pool water:
i level.
l
- d. The ADS system is actuated u?on coincident signals of the reactor-l vessellow waterlevel and any L?CI or CS pump running. If all signals-i are present the ADS valves will open after the ADS two minute timer l
runs out.- The time delay gives the HPCI system a chance to operate before blowdown occurs.
LPCI initiation automatically causes all RHR components to perform e.
their function under the LPCI node.
I
- 2. Remote manual i
ECCS pumps and valves and the. ADS can be actuated by remote manual means from the main control room.
~
l B. Motive Power
lE AC loads that can be supplied from the emergency diesel generators, as described in Section 3.5.
- 2. Most of the HPCI motor-operated valves are Class lE DC loads. The HPCI pump is supplied with steam from ~ main steam line_ A.L -
C. Other i
1, Lubricat!cn and cooling for the ECCS pumps are assumed to be supplied 1-locally.
i 2 ECCS pump room ventilation systems are cooled by the Service Water 1--
System (see Section 3.7). RHR pump seals are also cooled by SWS.
- 3. The RHR heat exchangers are cooled by the Service Water System (see-Section 3.7).
a 3.3.7 Section 3.3
References:
p 1.
Brunswick 1-& 2_ Updated Final Safety Analysis Report, Section 6.3.
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Table 3.3-1.
Brunswick 2 Emergency Core Cooling System Data Summary for Selected Com;,nents COMPCNENT ID COMP.
LOCATION POWER SOURCE VOLTAGE POWER SOURCE EMERG.
f TYPE LOCATION LOAD GRP_
?
CS-15A MOV NWCORHM EP-MCC-2XC 480 20HB AC/E3 CS-ISB MOV SWCORRM EP-MCC-2XD 480 20RB AC/E4 i
CS-1 A MOV NWCORHM EP-MCC-2XC 480 20RB AC/E3
.k CS-1B MOV SWCORiiM EP-MCC-2XD 480 20RB AC/E4 CS-4A MOV SORB EP-MCC-2XC 480 20RB AC/E3 I
CS-4 B MOV SORB EP-MCC-2XD 480 20HB AC/E4 i
CS-SA MOV 50RB EP-MCC-2XC 480 20RB.
AC/E3 i
CS-58 MOV.
50RB EP4ACC-2XD 480 20RB AC/E4 I
CS-P2A MDP NWCORRM-EP-BS-E3 4160 50DGSG3 AC/E3 CS-P2B MDP SWCORf1M EP-BS-E4 4160 50DGSG4 AC/E4 itPCI-1 MOV IIPCIRM EP-MCC-2XDA 125 20RB DC/28 l
HPCI-11 MOV RHRBRM EP-MCC-2XDA 125 20RB DC/2A w
HPCI-2 MOV RC -.
EP-MCC-2XD 480 20RB AC/E4 HPCI-3 MOV ECCSTNL EP-MCC-2XDA 125 20RS DC/2A HPCI-4 MOV HPCIRM EP-MCC-2XDA 125 20RB DC2B t
HPCI-4 MOV HPCIRM EP-MCC-2XDA 125 20RB DC/2A HPCI-41 MOV HPCIRM EP-MCC-2XDA 125 20RB-DC/2A HPCl42 MOV HPCIRM EP-MCC-2XDA 125 20RB DC/2A HPCI-59 MOV 11PCIRM EP-MCC-2XDA.
125 20RB DC/2A HPCI-6 MOV.
20RB EP-MCC-2XDA.
125 20RB DC/2A HPCI-7 MOV RHRARM.
EP-MCC-2XDA 125 20RB DC/2A-f HPCI-8 MOV RHRARM EP-MCC-2XDA 125 20RB DC/2A :
HPCI-lP TDP HPCIRM ilPCI-TSV8 HV.
HPCIRM -
oo RCS-32A MOV MSIVRM EP-MCC-2XA 480 20RB-AC/E3 RHR-10 MOV RHRARM EP-MCC-2XB 480 20RB AC/E4
[
RHR-10 MOV RHRARM-EP-MCC-2XB :
480 20RB AC/E4 RHR-15A MOV ECCSINL EP-MCC-2XA2 480 2DRB AC/E1
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Table 3.3-1.
Brunswick 2 Emergency Core Cooling. System Data Summary for Selected Components (Continued) t i
COMPONENT ID COMP.
LOCATION POWER SCURCE YO LT A G E POWER SOURCE EMERG.
}
TYPE LOCATION LOAD GRP.
RHR-158 MOV ECCSINL EP-MCC-2XB2 480 20HB AC/E2
)
RfiH-16A MOV 20RB EP-MCC-2XA 480 20HB AC/E3 RHR-16A MOV 20f18 EP-MCC-2XA 480 20RB AC/E3
)
fillR-168 MOV 20RB EP-MCC-2XB 480 20RB-AC/E4 RHR-168 MOV 20RB EP-MCC-2XB 480 20RB AC/E4 HHR-17A MOV.
RHRPIPECHASE EP-MCC-2XA2 480 20RB AC/E1 RHR-17B MOV RHRPIPECHASE EP-MCC-2XB2 480 20RB -
AC/E2 RHR-20A MOV RHRARM EP-MCC-2XA 480 20HB AC/E3 RHR-208 MOV RHflBilM EP-MCC-2XB 480 20RB AC/E4 flHR-21A MOV.
20RB EP-MCC-2XA 480 20HB AC/E3 RHR-21A MOV 20RB EP-MCC-2XA 480 20RB AC/E3 i
flHR-21B MOV-20RB-EP-MCC-2XB -
480 20RB AC/E4 g
RHR-21B MOV 20RB EP-MCC-2XB 480 20RB AC/E4 f
RHR-24A MOV' TOR EP-MCC-2XA 480 20RB AC/E3 i
RHR-24A MOV TOfl EP-MCC-2XA 480 20RB AC/E3 HHR-248 MOV IOR EP-MCC-2XB 480 20RB AC/E4 RHR-248 MOV TOR EP-MCC-2XB 480 20RB AC/E4 RHR-27A MOV.
TOR EP-MCC-2XA 480 20RB AC/E3 RHR-27A.
MOV-TOR EP-MCC-2XA 480 20RB AC/E3 RHR-278 MOV TOR EP-MCC-2XB -
480 20RB AC/E4 RHR-27B MOV TOR EP-MCC-2XB 480 20RB AC/E4 RHR-28A MOV TOR EP-MCC-2XA2 480 20RB AC/E1 RHR-28A MOV TOR EP-MCC-2XA2 480 20RB AC/E1 l
flHR-288 -
MOV TOR EP-MCC-2XB2 480 20RB AC/E2 h
RHR-288 MOV TOR EP-MCC-2XB2 480 20RB AC/E2 RHR-3A :
MOV RHRARM EP-MCC-2XA 480 20RB AC/E3
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RHR-38 MOV-RHHBRM EP-MCC-2XB.
480-20RB AC/E4 I
RHR-47A MOV RHRARM EP-MCC-2XA 480 20HB AC/E3
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Table 3.3-1.
Brunswick 2 Emergency Core Cooling System Data Summary for Selected Components (Continued) i COMPONENT ID COMP.
LOC ATION POWER SOURCE VO LT A G C POWER SOURCE EMERG.
I TYPE LOCATION LOAD'GRP.
RHR-478 MOV RHRBRM EP-MCC-2XB 480 20RB AC/E4 RHR-48A MOV RHRARM EP-MCC-2XA 480 20RB AC/E3 RHil-48A MOV RHRAllM EP-MCC-2XA 480 20RB AC/E3
'I RHR-48B MOV RHilBRM EP-MCC-2XB 480 20RB AC/E4
{
RHR488 MOV-RHRBRM EP-MCC-2XB 480 20RB AC/E4 f
RHR4A MOV flHRARM EPEUC-2XA 480 20B AC/E3 l
RHR-4B MOV RHRBilM EP-MCC-2XB 480 20RB AC/E4 s
RHR-4C -
MOV flHRARM EP-MCC-2XA 480 20RB AC/E3 RHR-4D MOV-RHRBitM EP-MCC-2XB 480-20RB AC/E4 RHR-6A MOV RHRAflM EP-MCC-2XA 480 20HB AC/E3 i
RHR-6B MOV RHRBRM EP-MCC-2XD -
480 20RB AC/E4
~illR-6C MOV HHRAHM EP-MCC-2XA 480 20HB AC/E3 f
y_
RHR-6D MOV RHRBRM.
EP-MCC-2XD 480 20RB AC/E4 RHR-HX2A HX-RHRARM-RHR-HX28..
HX-RHRBRM 4
RHR-P2A MDP RHRARM EP-BS-E3 4160 50DGSG3 AC/E3 RHR-P2B MDP RHRBRM EF-BS-E4 4160 50DGSG4 AC/E4 RHR-P2C -
MDP.
RHRARM.
EP-BS-E1 4160 SODGSG1 AC/E1 RHR-P2D MDP-RHRBRM-EP-BS-E2 4160 50DGSG2 AC/E2 a:
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Dmnswick 1 & 2 3.4 INSTRUh1ENTATION AND CONTROL (I&C) SYSTEh1S p) tU 3.4.1 Sntem Function The instrumentation and control systems consist of the Reactor Protection System (RPS), actuation logic and controls for various Engineered Safety Features (ESP) systems, and systems for the display of plant information to the operators. The RPS and ESF actuation systems monitor the reactor plant, and alert the operator to take corrective action before specified limits are exceeded. The RPS willinitiate an automatic reactor trip (scram) to rapidly shut down the reactor when plant conditions exceed one or more specified limits. The ESF actuation systems will automatically actuate selected safety systems based on the specific limits or combinations oflimits that are exceeded. A remote shutdown capability is provided to ensure that the reactor can be placed in a safe condition in the event that the main control room must be evacuated.
3.4.2 Svstem Definition The RPS includes sensor and transmitter units, logic units, and output trip relays that interface with the control circuits for components in the Control Rod Drive Hydraulic fystem (see Section 3.6). The ESF actuation systems include independent sensor and transmitter units, logic units, and relays that interface with the control circuits for the many different components that can be actuated.
A summary of data on selected I&C system components is presented in Table 3.4-1. The remote shutdown capability is provided by the remote shutdown panel in conjunction with normal automatic systems and local equipment controls.
3.4.3 Snipm Oneration fs A. RPS Q
The RPS has four input instrument channels and two output actuation trains.
The RPS monitors and automatically initiates a scram based on the following variables:
Neutron monitoring (APRht) system Neutron monitoring (IRht) system Neutron monitoring (SRht) system Neutron monitoring (LPRhi) system Neutron monitoring (RBhi) system Transversing in core probe (TIP) subsystem Reactor vessel high pressure Reactor vessellow water level Turbine stop valve closure Turbine control valve fast closure hiain steam line isolation valve closure Scram discharge volume high water level Primary containment high pressure hiain steam line high radiation hiain condenser low vacuum hiode switch in SHUTDOWN' Select rod insert in addition, a scram can be manually initiated. There are two scram buttons, one for trip logic A3 and one for trip logic B3. Depressing the A3 scram button pi deenergizes trip actuators A3 and opens corresponding contacts in trip actuator
- d logics A. A single trip system trip is the result. To effect a manual scram, the buttons for both trip logic A3 and trip logic B3 must be depressed. It is also 33 1/89
Brunswick 1 & 2 o
possible for the control room operator to scram the reactor by interrupting (v) power to the reactor protection system, This can be done by operating power supply breakers in the battery room.
To restore the reactor protection system to normal operation following any single trip system trip or scram, the trip actuators must be manually reset. Reset is possib:e only if the conditions that caused the trip or scram have been cleared and is accomplished by operating switches in the control room.
The Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) has been added as a means oflimitia the consequences of the unlikely occurrence of a failure to scram during an a;nticipated transient. The ATWS-RPT trips the reactor recirculation pumps on the occurrence of reactor vessel low low water level or reactor vessel high pressure.
B. ESF ESF actuation systems have up to four input instrument channels for each sensed parameter, and two output trains. In general, each train controls equipment powered from different Class lE electrical buses. The ESF systems that can be automatically actuated include the following (not a complete listing):
Emergency Core Cooling System HPCI CS LPC1/RHR ADS Standby power systems Service water system 5
b Various room cooling systems ECCS equipment room HVAC system e
Essential switchgear heat removal HVAC system Diesel generator HVAC system Service water pump room HVAC system Main control room HVAC system Details regarding ESF actuation logic are included in the system description for the actuated system.
C. Remote Shutdown The capability is provided to shutdown'the reactor and maintain it in a safe condition in the event the control room must be evacuated. In the event that the control room becomes uninhabitable, the reactor may be shut down by one of the following means:
The reactor may be shut down-by operating the scram buttons at the Reactor-Turbine Generator (RTG) board lbefore the Control Room is abandoned.
The reactor protection system will function automatically to shut down the reactor on reactor high pressure, low water level, or loss of power to the RPS MG sets. Two channels are provided so that even with the loss of one of these, the reactor would be shut down safely.
U 34 1/89
Brunswick 1 & 2 p}
The reactor may be shut down from the turbine trip lever at the turbine front g
standard. This will put the RPS into operation.
v The reactor may be shut down by cutting off power to the RPS MG sets at the motor control centers feeding them.
The reactor would then be maintained in a safe shutdown condition by operating the RCIC system in conjunction with the safety / relief valves. The RHR and SWS systems would also be required for suppression pool heat removal.
These system may be operated from outside the control room at the remote shutdown panel, at local panels such as motor control centers, or manually, as appropriate. The controls for remotely operated equipment are provided with key-locked" Normal Local" selector switches or other isolation features.
Sufficient instrumentation for monitoring the status of the reactor and primary containment and the operation of the RCIC and RHR systems is provided on.
the remote shutdown panel. This instrumentation is either independent of the main control room instrumentation or is provided with isolation features so that a malfunction of fire in the Control Building will not affect its operation (Reference 1).
3.4.4 Svstem Success Criterin A. PPS The RPS uses hindrance logic (normal = 1, trip = 0) in both the input and output logic. Therefore, a channel will be in a trip state when input signals are lost,
/O when control power is lost, or s.$n the channel is temporarily removed from V
service for testing or maintenance (i.e. the channel has a fail-safe failure mode).
A reactor scram will occur upon loss of control power to the RPS A reactor scram is implemented by the scram pilot valves in the control rod drive hydraulic system (see Section 3.6). Details of the RPS for Brunswick I and 2 have not been detemlined.
B. ESF Actuation Systems l
A single component usually receives a signal from only one ESF actuation system output train. Trains A and B must be available in order to automatically actuate their respective components. Actuation systems other than the RPS typically us. 'indrance input logic (normal = 1, trip = 0) and transmission output logic rmal = 0, trip = 1). In this case, an input channel will be in a trip state whet...put signals are lost, when control power is lost, or when the channel is temporarily removed from service for testing or maintenance (i.e, the l
channel has a fail safe failure mode). Control power is needed for the ESF actuation system output channels to send an actuation signal. Note that there may be some ESF actuation subsystems that utilize hindrance output logic. For these subsystems, loss of control power will cause system or component actuation, as is the case with the RPS. Details of the ESF actuation systems for Brunswick 1 and 2 have not been detennined.
C. Manually-Initiated Protective Actions When reasonable time is available, certain protective actions may be performed manually by plant personnel. The control room operators are capable of pi operating individual components using normal control circuitry, or operating d
.;roups of components by manually tripping the RPS or other actuation i
subsystem. The control room opera.o:s also may send qualified persons into 35-1/89
Brunswick 1 & 2 the plant to operate components locally or from some other remote control location (i.e., the remote shutdown panel or a motor control center). To make
.s these judgments, data on key plant parameters must be available to the operators.
3.4.5 Sunnort Systems and Interfaces A. Control Power
- 1. RPS The RPS is powered from the 120 VAC RPS system. Alternate power is available from an electrical bus that can receive standby electrical power.
Backup scram valves are powered from the 125 VDC system.
- 2. ESF Actuation Systems Control power sources for ESF actuation systems were not clearly defm' ed in the FSAR.
- 3. OperatorInstrumentation Operator instrumentation displays are powered from 120 VAC panels through transformers from the 480 VAC motor control centers E5, E6, E7, and E8.
3.4.6 Section 3.4 References
- 1. Bruns,vick 1 & 2 Updated Final Safety Analysis Report, Section 7.4.4.1.
{d Pv 36 1/89
Brunswick 1 & 2 lOV 3.5 ELECTRIC POWER SYSTEM i
3.5,1 System Function The electric power system supplies power to various equipment and systems needed for normal operation and/or response to accidents, The onsite Class 1E electric i
power system supports the operation of safety class systems and instrumentation needed to :
establish and maintain a safe shutdown plant condition following an accident, when the normal electric power sources are not available, 3.5.2.
System Definition I
The onsite Class IE electric power system consists of four independent'4160 and 480 VAC trains, denoted El, E2, E3, and E4. Each AC power division has a standby.
' diesel generator which serves as the AC power source when the normal source of offsite power is unavailable.
The DC system is a three;line system consisting of four 125/250.VDC divisions -
denoted 1 A, IB,2A,~ and 2B. These buses are each supplied by two battery chargers and i
two batteries. As shown in Figure 3.5 9, the DC buses are designed to supply either 125 or 250 VDC loads. This can be accomplished because one line is :" positive 125 volts, onc
{
is at neutral, and the third line is at negative 125 volts. One batter are connected between the positive-125 volt and neutral lines,y charger and one b The remaining battery charger and battery are connected between the neutral and negative 125 volt lines. The two--
. battery chargers, however, are supplied from the same 480 VAC busc
- The 120 VAC system consists of four instrument buses, supplied by the 480 -
VAC system through transformers.
Simplified one line diagrams of the electric power system are shown in Figures 3.5-1 to 3.511. A summary of data on-selected electric power system components is m
t.
presented in Table 3.51. A partial listing of electrical sources and loads is presented in p
Table 3.5-2, a
--3.5.3 S ystem' Onera tion Each Class IE 4160 VAC bus is provided with a normal offsite power su feeder and one standby diesel. generator. The normal power source for buses El, E2 E3,1 l
and E4 are buses ID, IC,2D, and 2C,'respectively.. Details of the station electric power i
system are shown in Figures 3.51 and 3.5-2.
l-The four standby diesel generators are started upon loss of offsite power at the -
i 4.16 kV bus, or a LOCA signal, or manual actuation, either locally.or in the control roomc Diesel generators 1,' 2,3, and 4 are connected to the 4160 VAC safeguard buses ElcE2,--
~
E3,'and E4, respectively. - Each diesel is connected to only one bus. In turn, each 4160 :
VAC safeguard bus supplies power to a 480 VAC load center bus through a transformerc
- Details of the 4160 and 480 VAC systems are shown in Figures 3.5 3 to 3.5 6c The Class IE DC system consists of four independent divisions. Each 125/250 VDC system is comprised of a set of two'125 V batteries, and two cha gers. Each battery
- - has a nominal manutacturer's eight hour rating of 1200 ampere hours to a minimum battery 1
terminal voltage of 210 V (for the 250 V nominal systems) but itlis not known how long the batteries can su 2XDB, in division'pply their required loads without recharging. Motor control center i
28, supplies power to RCIC system valves? Motor control-center 4
2XDA, in division 2A, supplied power to HPCI system valvesJDetails of the 125 VDC 1
and 120 VAC systems are shown in Ficu= 3.5-7,3.5 8, and 3.5 9.
~
-Instrument power is provided by four independent Class 1E 208/120 VAC.
buses. The instrument buses receive power from 480 VAC motor control centers (MCCs) through transformers, as shown in Figures 3.5-10 and 3.511.-
i 37' 1/89 I
1 Bmnswick 1 & 2 7
)
Redundant safeguards equipment such as motor driven pumps and motor U
operated valves are supplied by different buses or MCCs, For the purpose of discussion, -
this equipment has been grouped into " load groups", Load group "AC/El" contains components receiving electric power either directly ofindirectly from 4160 bus EL Load group "AC/E2" contains components powered either directly or indirectly from 4160 bus E2, Load group "AC/E3" contains components powered either directly or indirectly from 4160 bus E3, Load group "AC/E4" contains components powered either directly or indirectly from 4160-bus E4, - Components receiving DC power are assigned to-load-groups "I A", "1B", "2A" or "2B", based on the battery source, 3,5,4 System Success Criteria Basic system success criteria for mitigating transients and loss of-coolant accidents are defined'by front line systems, which then create demands on support systems, Electric power system success criteria are defined as follows, without'taking credit for cross ties that may exist between independent load groups:
Each Class lE DC load group is supplied initially from its respective battery (also needed for diesel startmg) i Each Class Ib AC load group is isolated from the non Class IE system and is supplied from its respective emerpency power source (i.e. diesel generator)
~
Power distribution paths to essential loads are intact Power to the battery chargers is restored before the batteries are exhausted 3,5,5 Comoonent Information.
A. Standby diesel generators 1,2,3,~ 4
- 1. Contmuous power rating: 3500 kW s
- 2. Rated voltage: 4160 VAC 3, Manufacturer: Nordbergh.anufacturing Company ~-
B, Station batteries l A, IB,2A,2B
- 1. Type: lead-calcium.
- 2. Rated voltage: 125 VDC -
3.5.6 Suonort Systems and' Interfaces A. Control Signals
- 1. Automatic L
The standby diesel generators are automatically started upan loss of voltagei l
on their associated 4160 VAC bus or on a LOCA signal,
- 2. Remote manual The diesel generators can be started, and many distribution circuit breakers can be operated from the main control room,
- 3. Localmanual l
The diesel generators can be started locally, B. Diesel Generator Auxiliary Systems l
The following auxiliaries are provided for each emergency diesel generator:
Cooling The service water system (see Section 3,7) provides for diesel cooling.
Fueling An independent day tank is provided for each diesel, Long term fuel tanks -
\\
are located underground near the diesel generator rooms.
38
.'l/89 z i
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C Lubrication Each diesel generator has a self-contained lubricatica system.
Starting An independer.t starting air accumulator is provided for euch diesel generator.
Control power Each diesel generator is dependent on 125 VDC power from a station battery for control power.
Diesel room ventilation fans provide room cooling during diesel operation.
C. Switchgear Room Ver.tilation Ventilation capabilities for the essential switchgear rooms could not be detemtined.
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39 1/89
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Table 3.5-1.
Brunswick 2 Electric Power System Data Summary for Selected Components COMPONENT ID COMP.
LOCATION POWER SOURCE VOLTAGE POWER SOURCE EMERG.
TYPE LOC ATION LOAD GRP.
EP-BC-2A-1 BC BAT 2A EP-MCG-2CA 480 2CSA DG/2A E P-BC-2A-2 BC BAT 2A EP-MCC-2CA 480 2CSA DC'2A EP-BC-26-1 BC BAT 2B EP-MCC-PCB 480
'2CSA DQ28 EP-BC-20 2 BC BAT 2B EP-MCC-2GB 480
.~'4A DG7A EP-BS-1 A BUS BAT 1A EP-BT-1 A 125 BATTA DC/1 A EP-BS-1 A BUS BAT 1A EP-BC-1 A-t 125 BATTA DCf1A EP-BS-1 A BUS BAT 1 A EP-BC-1 A-2 125 BAT 1A DQ1 A EP-BS-1 A BUS BATTA EP-BC-1 A-2 125 BAT 1A DOTA EP-BS-18 BUS BA T1B EF-B I-1B 125 BAT 1B DQ1B EP-BS-18 BUS BAT 1B EP-BC-1B-1 125 BAItB DQ18 EP-BS-2A BUS BAT 2A EP-BT-2A 125 BA T2A DC/2A EP-BS-2A BUS BAT 2A EP-BC-2A-1 125 BAI2A DC/2A EP-BS-2A BUS BA12A EP-BC-2A-2 125 BAT 2A DC/2A EP-BS-28
! BUS BAT 2B EP-B T-2B 125 BAT 2B DC/28 EP-BS-28 BUS BA118 EP-BC-2B-1 125 BAT 2B DC/28 EP-BS-28.
BUS BAT 2B EP-BG-28-2 125 BAI23 DC/20
~EP-BS-Et BUS 50DGSG1 EP-DG-1 4160 23DGA AC/E1 EP-BS-E2 BUS 50DGSG2 EP-DG-2 4160 23DGB AC/EZ EP-BS-E3 B'US 50DGSG3 EP-DG-3 4160 23DGC AC/E3 EP-BS-E4 BUS 50DGSG4 EP-DG-4 4160 23DGD AC/E4 EP-BS-E5 BUS 23DGSGS EP-TR-FB5 480 23DGSG5 AC/E1 EP-BS-E6 BUS 23DGSG6 EP-T R+BS 460 23DGSG6 AGE 2 EP-BS-E7 BUS 23DGSG7 p EP-TR-FBI 480 23DG5G7 AC/E3 EP-BS-E8 BUS 23DGSGB EP-T R-FB2 480 23DGSG3 AGE 4 EP-BS1B BUS BAT 1B EP-BC-18-2 125 BAT 1B DQ18 EP-BT-1 A BAT BATTA 125 DQ1A EP-BT-18 BAT BAT 1B 125 DQtB EP-8 T-2A BAT UAT2A 125 DC/2A
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Table 3.5-1.
Brunswick 2 Electric Power System Data Summary for Selected Components (Continued)
COMPONENT ID COMP.
LOCATION POWER SOURCE VOLTA GE POWER SOURCE EMERG TYPE LOC A TION LOAD GRP.
EP-B T-28 BAT BAT 2B 125 DG/28 EP-DG-1 DG 23DGA AC/E1 EP-DG-2 DG 23DGB ACiE2 2
EP-DG-3 DG 23DGC AC/E3 EP-DG-4 DG 23DGD AGE 4 EP-MCC-1XDA MCC 20RB EP-BS-1 A 125 BAT 1A DQ1 A EP-MCC-1XDB MCC 20RB EP-BS-1 B 125 BAT 1B DQ1B EP-MCC-2CA MCC 2CSA EP-BS-E7 480 23DGSG7 AC/E3 EP-MCC-2CB UCC 2CSA EP-BS-E8 480 23DGSGB AGE 4 EP-MCC-2PA MCC 2 Chi 2A E P-BS-E7 480
' 23DGSG7 AGE 3 EP-MCC-2PB MCC 20SW2B EP-BS-E8 480 23DGSG8 AC/E4 EP-MCC-2XA MCC 20RB EP-BS-E7 480 23DGSG7 ACTE 3 EP-MCC-2XA2 MCC 20RB EP-BS-E5 480 23DGSGS AC/E1 EP-MCC-2XB UCC 20RB EP-BS " '
480 23DGSG8 AGE 4 EP-MCC-2XB2 MCC 20RB EP-BS-EG 480 23DGSG6 AC/E2 EP-MCC-2XC MCC 20RB EP-BS-E7 483 230GSG7 AC/E3 EP-MCC-2XD MCC 20RB EP-BS-EB 480 23DGSG8 ACiE4 EP-MCC-2XDA MCC 20RB EP-BS-2A 125 BAT 2A DC/2A EP-MCC-2XDB MCC 20RB EP-BS-28 125 BAT 2B DC/2B EP-MCC-OGA MCC 23DGA EP-BS-E5 480 23DGSGS AC/E1 EP-MCC-DGB MCC.
23DGB EP-BS-E6 480 23DGSG6 AC/E2 EP-MCC-DGC MCC 23DGC EP-BS-E7 480 23DGSG7 AC/E3
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TADLE 3.5 2, PARTIAL LISTING OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICK 2
~
f vi, t n voLiAGE EMERa POWER SOUME LOAD EOAD COMP COMPONEN1 SOLM E ECAD GRP LOCATION SYSTEM COMPONENT ID TYPE EOOATION E b bC-1 Al 120 DC/1 A BAT 1A EP EP bS 1A EvS bATIA t hbC 1 A i lit DC/1 A bA11A EP EP BS 1A BUS DAT1A EL & 1A-2 12b DC/1A BAT 1A EP EP BS 1A BUS BAT 1A EP bC 101 12L DC/16 BA118 EP EP BS1B BUS BATIB EP bC 1b 2 12b DC/1 b BA11B EP EP~BStB BUS BATIB E P BC i A-1 12L DC 2A BA12A EP EP BS 2A BUS BAT 2A EP bC 2A 2 126 DC/2A BAT 2A EP EP BS-2A BUS BAl2A EP bC-2D-1 126 DC/2B BATED EP EP BS-20 BUS BA11B EPhib 2 li!
DC/2B BATEB EP EP BS28 BUS BAT 2B ik bS1A lib DCt1 A bATIA EP EP MCC 1XDA MCC 20RB EP bS 10 126 DC/1B BA110 EP EP MCC 1 ADB MCC 20RB EP bb ; A 12h DC/iA BAT 2A EP EP MCC 2XDA MCC 20RB E P L S i.,
125 DC/26 BAT 2B EP EP MCC 2XDB MCC 20RB EP bS-E i 4160 AC<El SODGSG)
ECCS RHR-P2C MDP RHRARM E P-bb-E 1 4100 AC/E1 50DGSG1 EP EP lR FB5 TRAN 23DGSG6 EP BS Ei 4160 AC<Et 50DGSG1 SWS CSW P2C MDP SWBLDG EP bdE2 4160 ACiE 2 60DGSG2 ECCS RHR P2D MDP RHRBRM Ebb 5 E2 4160 ACE LODGSGP EP EP-TR Fb6 TRAN 23DGSG6
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$WS CSW P2A MDP SWBLDG EP BS E3 4160 AC/E3
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MDP SWCORRM EPBSEd 4160 AC/E4 50DGSG4 ECCS RHR-P2B MDP RHRBRM EP bS E4 4100 A C. E 4
$0DGSG4 EP EP1RFB2 TRAN 23DGSG8 EP BSE4 4160 AC.E4 50DGSG4 SWS CSW-P2B MDP SWBLDG EP BS E4 4160 AC/E4 50DGSG4 SWS NSW P2B MDP SWBLDG EP BS E5 460 AC/E1 23DGSG5 EP EP-MCC-2AA2 M00 20RB 4
EP bS E5 460 AC/E1 23DGSGS EP EP MCC DGA MCC 23DGA (v\\j EP BS-E0 460 ACcE2 23DGSG6 -
EP EP MCC-2XB2 MCC 20RB 53 1/89
- - _ _..... - _.. - -.. - ~ -
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TABLE 3.5 2.
PARTIAL LISTING OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICK 2 (CONTINUED) foaE4 v0LTAGE EMERG POWER SOURCE LOAD LOAD COMP COMPoh ENT 1
SOUA0E LOAD GRP EOCATION SYSTE M COMPONENT ID TYPE SCATION kk bS Et 460 AC/E2 23DGSG6 EP EP-MCC DGB EC 230GB EF bS E7 460 ACsE3 23DGSG7 EP E P-MCG-2C A EC 2CSA LO 65 E7 400 AC/E3 230GSG7 EP EP MCC 2PA EC 20SW2A EP-bS L7 460 ACiE3 23DGSG7 EP EP-MCC-2AA.
MCC.
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EP bb E6 480 AC/E4 230G%8 EP EP MCC 2AB MCC 20RB EP bS EL 460 AC/E4 23DGSGB EP EP-MCC 2AD MCC 20RB EP bbi6 460 AciE4 230GSG6 EP EP-McC DGD EC 230GD EP BS-E 6 460 AC<E4 23DGSGB RC40 RCS 320 MOV MSivRM TEJi31 1 A 12b DC/1 A BAT 1A EP EP BS-1 A BUS BATIA EP BT 1B 125 9C/1B BAT 1B EP EP BS 1B BUS BATIB EPbl2A 126 DC/2A BAl2A EP EP BS-2A BUS BAT 2A
EP DG 1 4160 AC/E l 23DGA EP EP BS E1.
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TADLE 3.5 2.
PARTIAL LISTING OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICK 2 (CONTINUED)
POMR VOLTAGE EMERG POWER SOUROE LOAD LOAD COMP COMPONENT LOURCE LOAD GRP LOCATION SYSTEM COMPONENT ID TYPE LOCATION EP MeCU B 480 AciE 4 t0SW2B SWS SWS 20 MOV bwBLDG EP MGC 2AA 460 AC/E3 2040 ECCS RC5 32A MOV MSIVRM EP-MOC 2AA 400 AC/E3 20RB ECCS RHR 16A MOV 204)B EP+MCC 2AA 480 AC/E3 20RB ECCS RMR 16A MOV 20RB EP MCC 2AA 460 AC/E 3 20RB ECCS RHR 20A MOV RHRARM EP MCC ZAA 450 AC/E3 20RB ECCS RHR 21 A MOV 20RB EP MCC 2AA 460 AC/E3 -
20RB ECCS--
RHR 21 A MOV 20RB FUbC-2AA 460 AC/E3 20RB ECCS RHR 24A MOV TOR EP MCC.2AA 400 AC/E3 204D ECCS RHR 24A MOV-LOR E P MCC 2 AA -
4BD AC/E3 20RB ECCS RHR 27A MOV TOR EP-MCC 2AA 460
~
AC<E3 204B ECCS RHR 27A MOV TOR EP-MCC 2AA 460 AC/E3 20RB ECCS RHR 3A MOV RHRARM EP MCC 24A 460 AC/E3 20RB ECCS RHR 47A MOV RHRARd EP MCC 2AA 460 AC/ E3 204B ECCS RHR 48A MOV RHRARM EP MCC 2AA 480 AC/E3 20RB -
ECCS RHR 48A MOV RHRARM EP MCG-2AA 460 AC/E3 200 ECCS RHR4A MOV RHRARM EP MCC 2AA 480 AC/E3 20RB ECCS RHR4C MOV RHRARM EP MCC 24A 460 AC< E3 20RB ECCS RHR 6A
'AOV RHRARM i
20RB ECCS RHR4C 7V RHRARM EP MCC 2AA 480 AC/E3 20RB RCS RHR** ~
"' OV RC EP MCC 2AA2 480 AC/E1 2DRB ECCS R' 415A MOV ECCSTNL i
EP MCC 2AA2 460 AC/E l 20RB ECCS.
EP MCC 2AA2 460 AC/E 1 20RB ECCS RHR 28A.
MOV TOR-EP MCC 2AA2 460 AC/E 1 20RB ECCS RHR 28A MOV TOR EP MCC-2XB 480 AC/E4 20RD ECCS RHR 10 MOV RHRARM-EP MOC 2AB 480 AC/E4 20RB ECCS-RHR 10 MOV RHRARM EP MCC-2AB 460 AC. E 4 20RB ECCS RHR 16B MOV 20RB 1
20RB ECCS RHR 16B MOV 20RB EP MCC 2AB 480 AC/E4 20RB ECCS RHR 20B-MOV RHRBRM EP MCC 2xB -
480 AC/E4 20RB ECCS RHR 210 MOV.
20RB 55 1/89 ' e
~........ _..
TABLE 33 2.
PARTIAL LISTING OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICX 2 (CONTINUED) h.inim m iACE EMEk3 POW E R SOURCE LOAD LOAD COMP COMPONENT SOUROE LOAD GRD LOCATION SYSTE M COMPONENT ID TYPE LOCATION EPhCOD 460 ACiE 4 20Rb ECCS RHR 24B MOV TOR EP Mec 2 Ab 4to ACrk4 20RD ECCS RHR 24B MOV 104 EP MCC 2AD 400 ACE 4 20RB ECCS RHR 27B MOV TOR EP MCC 2Ab 460 AC4E4 20Ttb ECCS RHR 27b MOV TOR EhMCC 2AB 460 AC/E4 20Rtl ECCS RHR 3B MOV RHRBRM LPMCC4AD 400 AC<E4 20RD ECCS RMR 47B MOV RHRGRM EP McC4Ai) 400 AC/E4 20RB ECCS RHR 4BB MOV RHkBRM EP MCC4AB 460 AC/E 4 20Rb ECCS RHR 4BB MOV RHRBRM EP M^,04AD 4eC AC'E4 20RB ECCS RHR-4B MOV RHRBRM EFMCC4AD dBU AC<E d 20RD ECCS RHR4D MOV RHRORM EP MCC4Ab2 480 AC,E2 20Rb ECCS RHR 160 MOV ECCSINL E h McC4 Ab2 460 AC< E 2 20Rb ECCS RHR 17B MOV RHRP PECHASE EkMcc 2Ab2 460 AC<E2 20RB ECCS RHR 260 MOV LOR EP-MCC-2 Ab2 460 AC E2 20kB ECCS RHR 26B MOV TOR EP MCC 2 AC 400 AC/E3 20HB ECCS CS-15A MOV NWCORRM EP MCC 2 AC 460 AC/E3 20kB ECCS CS-1 A MOV NWCORRM EP MCC4 AC 480 AC/E 3 20RB ECCS C S-4 A MOV bORB EP-MCC 2 AC 460 AC,E 3 20RB ECCS CS bA MOV 60RB E P-MCC4 AC 4ec AC/E3 20RB RllC RCiG-7 MOV RC EPMCC-2AC 400 AC/E3 20RB RCS-RCIC7 MOV RC EP-MCC 2 AC 430 AC/E3 20RB RCS RCS1 MOV RC EP-MCC 2AC 460 ACsE3 20RB RCS RCS 16 MOV RC EP-MCC 2AD 480 AC/E4 20RB ECCS CS 168 MOV SWCORRM EP MCC-2 ND 460 AC/E4 20RB ECCS CS 1B MOV SWCORRM EP MCC4AD 480 ACiE4 20RB ECCS CS-4 B MOV 50RB EP MCC 2AD 480 AC/E 4 20RB LCCS CS-6B MOV
$0RB EP MCC 2AD 480 AC/E4 20RB ECCS HPCI-2 MOV RC EP-McC4AD 460 AC/E4 20RB ECCS RHR4B MOV RHRBRM EP.MCC-2 AD 460 AC/E4 20RB ECCS RHRfD MOV RHRB4M EP MCC 2AD 460 AC<E4 20RB RCS HPCI-2 MOV RC EP-MCC-2ADA 125 DC/2B 20RB ECCS HPCI1 MOV HPCIRM a
56 1/89
-. -.. -. ~ _..
=-.-
TADLE 3.5 2.
PARTIAL LISTil4G OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICK 2 (CONTINUED)
POA%
v0.1 AGE E ME RG PO AE R SOVRCE LOAD LOAD COMP COMPONENT SDJR2E LOAD GRP LOCATION SYSTE M COMPONENT ID TYPE LOCATION E u;L'iALA lib DCc2A 20Rb ECCS HPCI 11 MOV RMRbkM EF M004ADA IIL DC4A 20Rb ECCS HPC64 MOV ' ECCSINL EP-MaC4ADA 126 DC/2b 20RD ECCS HPCI4 MOV HPCIRM EP MLC4ADA 126 DC/2A 20RB ECCS HPCl4 HOV HPCiRM EP,MGC4ADA 12t DC4A 2DRB
- nCS P"C
- 4, eloif HPCIRM Ek MGC4ADA lib DC2A 20RP ECCS HPCI-42 MOV HPC4RM EP MCC4ADA 126 DC4A 20f.B ECCS HPCI-b9 MOV HPCIRM EP MCG ?ADA 12b DC/2A 20k0 ECCS HPCl4 MOV 20RB IPMLC4ADA 126 DC/2A 20kb ECCS HPCI7 MOV RHRARM E F.M,04 ADA 12b DC.2 A 20RD ECCS HPCl4 MOV RhRARM EP M;CGADA til D C. i A 20RD ACS hPCI-3 MOV ECCSINL EP MvC4ALb 12t DC2b 20RB RCIC RCIC to MOV RHkBRM EP MLC4 ADb ist DC40 20RB RCIC RCIC 11 MOV RHRBRM EPM;C4 ADD 126 DC/IB 20R0 RCiG RCIC 12 MOV RHRBRM EbMeC4ALc 12b 0C/2B 20RD ACIC RCIC13 MOV 20RB EP MCC4ADb 126 DC/IB 20RA ACIC RCIC-19 MOV RHRBRM EbMeC4ARb 126 DC ib 20RB RCIC RCIC42 MOV RHRBRM EPMJC2ADB 126 DC4B 2040 RCIC RCIC 29 MOV RHRuRM EP MCC4 ADb 12b DC/28 20RB RCIC RCIC41 MOV RHABRM l.P MCC4 ADb 126 DC/2 B 20RB RCIC RCIC 45 MOV RHRBRM TUMOC4ADB 126 DC/20 20RB R0lO RCIC 46 MOV RHRBRM EP M004 ADD 126 DC/2B 20RB
'ICIC RCIC4 MOV ECCSlNL EP MOC4ADb 126 DCJD 20RB A010 RCIC TTV8 MOV RHRBRM EP MCC4ADb 126 DC4B 20RB RCS RCIC-8 MOV.
ECCSTNL LP MOC4 AEB 126 DC/20 20RB RCS RCS-19 MOV MSIVRM EhMOC4ADE 126 DC40 20RB RCS ACS4 MOV RWCUPRM EP MCC4ADB 126 DC/2B 20RB RCS RHR4 MOV ECCSINL Ek MOC DGD 4b0 AC/E4 23DGD SWS SWS 256 MOV V255RM EP TR461 400 AC/E3 23DGSG7 EP EP DS E7 BUS 23DGSG7 TP 1R4b2 dbo AC/E 4 23DGSGS EP EP BS-E8 BUS 23DGSG8
(
EP1R4Bb 460 AC/E1 23DGSG5 EP EP BS-E5 BUS 23DGSG5
$7 1/89
T.TDL E 3.5 2.
PARTIAL LISTING OF ELECTRICAL SOURCES AND LOADS AT BRUNSWICK 2 (CONTINUED) f 0.'. L 4 vo$TAGE EME4G POWE R 60J4CE LOAD LOAD COMP COMPONENT LOURCE LOAD GRP LOCATION SYSTEM COMPONENT 10 TYPE LOCATION
o v
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Brunswick 1 & 2 3.6 CONTROL ROD DRIVE ilYDRAULIC SYSTEM (CRDilS) 3.6.1 Sntem Function The CRDHS supplies pressurized water to operate and cool the control rod drive mechanisms during normal operation. This system implements a scram command from the reactor protection system (RPS) and drives control rods rapidly into the reactor.
The CRDHS also can provide makeup water to the RCS.
3.6.2 System i erinition The CRDils consists of two high head, low flow CRD supply pumps, piping, filters, control valves, one hydraulic control unit for each control rod drive mechanism, and instrumentation. Water is supplied from the condensate treatment system or the condensate storage tanks. The CRDHS also includes scram valves, scram accumulators, and a scram discharge volume.
Details of the scram ponion of a typical BWR CRDHS is shown in Figure 3.6 1.
3.6.3 Sntem Onerntion During normal operation the CRDHS pumps provide a constant flow for drive mechanism cooling and system pressure stabilization. Excess water not used for cooling is discharged to the RCS. Control rods are driven in or out by the coordinated operation of the direction control valves, insertion speed is controlled by flow through the insert speed contral valve. Rod motion may be either stepped or continuous.
A reactor scram is implemented by pneumatic scram valves in the CRDilS. An inlet scram valve opens to align the insert side of each control rod drive mechanista (CRDM) to the scram accumulator. An outlet scram valve opens to vent the opposite side
.V of each CRDM to the scram discharge volume. This coordinated action results in rapid insertion of control rods into the reactor.
The control rod drive accumul; tors are necessary to scram the control rods within the required time. It should be noted that each drive has an internal ball check valve which allows reactor pressure to be admitted under the drive piston. If reactor pressure exceeds the supply pressure at the drive, the ball check valve ensures rod insertion in the event that the scram accumulator is not charged or the inlet scram valve fails to open. The insertion time, however, will be slower than the scram time with a properly functioning scram system.
Although not intended as a makeup system, the CRDHS can provide a source of cooling water to the RCS during vessel isolation. In BWR/4 plants, RCS makeun at high pressure is performed by the RCIC (see Section 3.2) and HPCI (see Section 3.3) systems. The maximum RCS makeup rate of the CRDHS is about 200 gpm with both pumps operating (Ref.1),
- 3. 6..I Svctem Success Criterin For the scram function to be accomplished, the following actions must occur in the CRDllS:
A scram signal must be transmitted by the RPS to the actuated devices (i.e.,
pilot valves)in the PdDHS.
The pneumatic inlet scram valve and outlet scram valve must open in the hydraulic control units (HCUs) for the individual control rod drives. This is accomplished by venting the instrument air supply to each valve as follows:
Both scram pilot valves in each HCU must be deenergized, or Either backup scram pilot valve must be energized, l
59.
1/89 t-
Brunswick 1 & 2 O
A hydraulle vent path to the scram discharge volume must be available and A high pressure water source must be available from the scram accumulator in each IICU.
sufficient collection volume must exist in the scram discharge volume.
core (specific number needed is not known). ponds and insert into the reac A specified number of control rods must res i
l l
3.6.5 Comonnent information A. Control rod drive pumps (2)
- 1. Rated capacity: 1007c (for control rod drive function)
- 2. Type: centrifugal i
B. Condensate Sterage Tank
- 1. Capacity: 1,000,000 gal C. Scram Accumulator
- 1. Nomial pressure: 14001500 psig D. Scram Discharge Volume
- 1. Normal pressure: Atmospheric 3.6.6 Sunnnri Systems nnd interfaces A. Control Signals
- 1. Automatic O
The RPS transmits scram comms. ids to solenoid pilot valves which control V
the pneum:stic scram valves
- 2. Remote Manual
- a. A reactor scram can be initiated manually from the control room
- b. The CRDHS can be operated manually from the control room to insert and withdraw rods, or to inject water into the RCS B. Motive Power
- 1. The control rod drive pumps are Class lE AC loads that can be supplied from the emergency diesel generator as described in Section 3.5.
3.6.7 Section 3.6 References
- 1. Harrington, R.M., and Ott LJ., "The Effect of Small Capacity, High Pressure Injection Systems on TQUV Sequences at Browns Ferry Unit One,"
NUREG/CR 3179, Oak Ridge National Laboratory, September 1983.
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3,7 SERYlCE WATER SYSTEM (SWS) AND RESIDUAL IICAT V
REMOYAL SERYlCE WATER SYSTEM (RSWS) 3,7,1 Sutem Function The Service Water System (SWS) for each unit is designed to operate under normal operating, shutdown, and design basis accident conditions. For normal and shutdown conditions, the SWS provides cooling water to the Reactor Building and Turbine Building Closed Cooling Water systems and supplies bearing lubrication water to the circulatmp water pumps. During design basis accident conditions, the SWS serves as a heat sink for the general cooling tequirements of the following components:
Emergancy diesel engine coolers RHR pump seal cooling heat exchangers Fan cooling units for the Core Spray and RHR pump rooms AOG System precoolers Residual Heat Removal Senice Water System (RSWS)
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Reactor Building Closed Cooling Water (RBCCW) system The RSWS operates in tandem with the SWS and supplies cooling water to the RilR heat exchangers. The RSWS system also serves as a backup for core flooding, 3,7,2 Sutem Definition The SWS is subdivided into two major portions, one basically for nuclear and i
vital loads and tne other normally for conventional loads in the Turbine Building. The two portions of the system are nonnally operated independent 1v, each consisting of a group of
,p service water pumps, parallel loads, and interconnecting headers,. Suitable cross-connecting valves and piping are provided to permit use of the conventional system as a
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backu a supply for reactor building cooling loads Backup for diesel generator cooling is provited by the nuclear headers of each unit or by cross connecting conventional heade' pumps to the nuclear header, The RSWS system consists of two trains, each with two pumps taking a suction on an SWS supply header and delivering cooling water to one RHR heat exchanger, The SWS and RSWS system share common discharge headers, Simplified drawings of the SWS and RSWS are shown in Figures 3.71 and 3,7 2. A summary of data on selected SWS and RSWS components is presented in Table 3,71, 3,7,3 System Onerntion Service water is supplied by five vertical intake structure that take suction from the intake canal. pumps installed in the senice water Under normal operating conditions, two pmps (one operath.g, one spare) provide water to the nuclear header, while the remaming three pumps (two operating, one spare) furnish water to the conventional header, Components normally supplied from the service water nuclear header are the Reactor 1
Building Closed Coohng A ater heat exchangers, RHR "B" heat exchanger, all RHR pump seal cooling heat exchangers, all fan cooling units for the pump rooms in the Reactor Building (core spray and RHR pumps rooms), RHR service water pumps "B" and "D" and their motor cochng, and standby diesel generator cooling. The senice water conventional header normally supplies cooling water to the Turbine Building Closed Cooling Water heat exchangers, lubncating water for the circulating water pump beatings and fill water for the circulating water system. When required, the conventional header also provides cooling water to the RHR "A" heat exchanger and the RHR service water pumps "A" and "C" and their motor cooling. These parallel heat loads can be cross connected between each of the i
service water headers by means of nonnally closed cross-connect valves.
62 1/89
(
o Brunswick 1 & 2 Doth service water headers supply their respective parallel loads in the Reactor Building in a similar piping arrangement. Service water flow to each component is initiated or terminated by means of motor operated valves which are open prior to placing that com aonent in ;ervice. Locked open manual valves are also installed in each su aply header.
Coo:ing water is delivered to the tube side of the RHR heat exchangers from e tier service water header via RHR service water pumps. Each of the two RHR heat exchangers is served by two RHR service water pumps, which under normal conditions provides sufficient pressure to assure that there will be no leakage of reactor coolant from the shell side of the RHR heat exchangers into the service water. Each pump delivers 4,000 ;pm of water from a common service water suction header provided with a nomially closec cross-connect isolation valve between the two pairs of pumps. This valves provides for separability, so the tailure of a component in one RHR senice water loop wiu not affect the function of the other, 3.7.4 Sv< tem Success Criterin
- 1. The success enteria for the SWS are that one of five senice water pumps must operate and the flow path to the heat load must be open.
- 2. The success criteria for the RSWS is: (a) sufficient coolant from the SWS must be supplied (b) two of four RSWS pumps must supply water to the same heat exchanger, and (c) the flow path to the heat load and discharge canal must be open.
3.7.5 Comnonent information A. N1 clear llender Service Water System Pumps P2A and P2B
- 1. P :at flow: 8000 gpm @ l15 ft, head (50 psid)
- 2. Rated capacity: 10We
- 3. Type: two stage, vertical turbine, wet pit, centrifugal B. Conventional licader Service Water System Pumps P2A, P2B, P2C
- 1. Rated flow: 8000 ppm @ l15 ft. head (50 psid)
- 3. Type: two stage, vertical turbine, wet pit, centrifugal C. Residual IIcat Removal Service Water Pump 47 psid) s P2A, P2B, P2C, P2D
- 1. Rated flow: 4000 ppm @ 570 ft, head (
- 2. Rated capacity: 509c 3.7.6 Sunnnrt Systems nnd interfaces A. Control Signals
- 1. Automatic i
The conventional service water header and the nuclear service water header pumps are connected for automatic operation (Ref.1. Section 9.2.1.5). A decrease of head pressure below the set point will start the spare pump in the respective header. Each unit is provided with an opposite unit starting signal to the nuclear service water pumps in the event of a LOCA or loss of offsite power.
J 63 1/89
l Brunswick 1 & 2 I
- 2. Remote manual
- a. The SWS pumps can be actuated by remote manual means from the i
control room. SWS pumps A and loop A valves can also be controlled from the remote shutdown panel.
l
- b. The RilRSW system is actuated manually.
B. Motive Power The SWS and RiiRSW pumps and motor operated valves are Class lE AC loads that can be supplied from ti ? standby diesel generators as described in Section 3.5.
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Table 3.7-1.
Brunswick 2 Service Water System Data Summary l
for Selected Components COMPONENT ID COMP.
LOCATION POWER SOURCE VOLTAGE POWER SOURCE EMERG.
TYPE LOCATION LOAD GRP_
CSW-P2A MDP SWBLOG EP-BS-E3 4160 50DGSG3 AC/03 CSW-P2B MDP SWBLDG EP-8S-E4 4160 50tX3S"A ACTE 4 CSW V2C MDP SWBLDG E P-BS-Et 4160 500GSG1 AC/E1 i
NSW-P2A M9P SWBLDG EP-BS-E3 4160 50DGSG3 AC/E3 NSW-P2B MDP SWBLDG EP-BS-E4 4160 50DGSG4 AC/E4 SWS-13 MOV SWBLOG EP-MCC-2PA 460 20SW2A
. AC/E3 SWS-14 MOV SWBLOG EP-MCC-2PA l480 20SW2A AC/E3 SWS-15 MOV SWBLDG EP-MCC-2PB 480 20SW2B AC/E4 SWS-16 MOV SWBLDG EP-MCC-2PB 480 20SW2B AC/E4 SWS-17 MOV SWBLOG EP-MCC-1 PA 480 10SW1 A AC/E1 SWS-18 MOV SWBLD(2 EP-MCC-1PA 480 10GW1A AC/E1
^
SWS-19 MOV SWBLDG EP-MCG-2PA 480 20SW2A AC/E3 g
SWS-20 MOV SWBLDG EP-MCC-2PB 480 20SW2B AC/E4 SWS-255 MOV V255HM EP-MCC-DGD 480 23DGD AC/E4 i
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l
_... _ ~ _.. _
1 Brunswick 1 & 2
- 4. PLANT INIT 'MATIG N' 4.1 SITE AND flUILDING
SUMMARY
l The Brunswick site is located in the southeastern portion of North Carolina in Brunswick Count Cape Fear River. y, approximately two miles north of Southport and two miles site and Unit 2 at the south end. The site comprises a pproximately 1200 acres. The site is owned by Carolir a Power & Light (CP&L) and the North Carolina East Municipal Power Agency. The city J Wilmington, North Carolina is 16 miles northeast of the Brunswick site. A general view of the site is shown in Figure 41 (from Ref.1) and a more detailed site plan is shown in Figure 4 2.
The two reactor buildings for Units 1 and 2 are separated by control and radwaste buildings that are shared by the two units. A turbine building, containing the Unit -
- 1. and 2 main turbines and balance of plant systems,is adjacent to the west sides of the Unit I and 2 reactor buildings and the control building.
At each unit, the containment is surrounded by the reactor building. The HPCI, RCIC, core spray, RHR and reactor water cleanup systems, and the CRD hydraulic control units are located on various elevations of the reactor building. The spent fuel storage pool is on the 117 foot elevation and the remote shutdawn panel is on the 20 foot elevation of the reactor building. An ec uipment hatch and a personnel air lock are located on the 20 foot elevation of the reactor bui. ding.
The control building, located between the two reactor buildings, contains the control room, cable spreading rooms, battery rooms, and control building HVAC room.
The control rooms for Units 1 and 2 are in a common area on the 50 foot elevation of the control building' diesel generator building for Units 1 and 2 is located to the c The (O) both units are located in the diesel building. The long term fuel oil storage tanks are in a radwaste building. Four diesel generators and 4160 VAC Class IE switchgear serving below grade structure adjacent to the diesel generator building.
The service water pum1 house and the circulating water pumphouse are located on the intake canal on the east sic e of the site. Each pumphouse contains the respective pumps for both Units 1 and 2, The intake canal draws water from the Cape Fear River which is located to the cast of the site. The discharge " canal" exits underground on the west side of the site and discharges to the Atlantic Ocean; The transformer yard and switchyard for both Units 1 and 2 are located to the west of the turbine building.
One condensate storage tank (CST) is provided for cach unit. The Unit 1 CST is east of the Unit I reactor buildmg, near the norddeast corner of the diesel building. The Unit 2 CST is cast of the Unit 2 reactor building. Fire water and demineralized water storage tanks for both units are located at the north east corner of the Unit I reactor building.
Personnel and vehicie access to the protected area is through an access control point adjacent to the administration building at the south west corner of the site. CP&L.
L owns and operates a rail line serving the site. Rail access is provided at two points on the L
nonh side of the site.
l 4.2 FACILITY LAYOUT DRAWINGS l
A section view of the of Brunswick reactor building (typical of Units 1 and 2)is :
i shown in Figure ?J. Simplified layout drawings for both Units 1 and 2 are presented in l
Figures 4 4 to 418. Major rooms, stairwi.ys, elevators, and doorways are slown in the i __
simplified layout drawings, however, many interior walls have been omitted for clarity, Labels printed in uppercase correspond to the location codes listed in Table 4-1 and used in 4
the component data listlags and system drawings in Section 3. Some additionallabels are-included for information and are printed in lowercase type.
F 68'
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. I._ 4 Brunswick 1 & 2 A listing of components by location is presented in Table 4 2. Components f
(
included in Table 4 2 are those found in the system data tables in Section 3, therefore this table is only a partial listing of the components and equipment that are located in a panicular nom or arenf the plant.
4.3 SECTION 4 REFERENCES
- 1. lieddleson, F.A., " Design Dat and Safety Features of Commercial Nuclear
't Power Plants.", ORNL NSIC 5.i, Volume II, Oak Ridge National Laboratory, J
Nuclear Safety Infomiation Center, January 1972.
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L 1
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Definition of Brunswick 2 Building and Table 4-1.
Location. Codes
.Ciu!n Descrintions 1.
I17RB Spent fuel pool operating tloor,' located on'the 117' elevation of the Reactor Building 2.
20RB 20' elevation of the Reactor Building 3,
2CSA Unit 2 Cable Spreading Area, located on the 23' elevation of the Control Building 4.
BATIA Unit 1 Battery Room, Division l A, located on the 23' elevation of the Control Building I
5.
BATIB Unit 1 Battery Room, Division IB, located on the 23' elevation -
of the Control Building 6.
BAT 2A Unit 2 Battery Room, Division 2A, located on the 23' elevation of the Control Building 7
BAT 2B Unit 2 Battery Room, Division 2B, located on the 23' elevation of the Control Building O
8.
CR Control room, located on the +49' elevation of the Control Building cmmon to both units-9.
CST Condensate Storage Tank, located in the Yard Area - for each unit the CST is located east of the respective Reactor Building
- 10. ECCSTNL Emergency Core Cooling System Pipe Tunnel, located on + 20' elevation of the Reactor Building
- 11. HPCIRM High Pressure Coolant Injection Pump Room, located on the
-17' elevation of the Reactor Building
- 12. MSIVRM Main Steam Isolation Valve Room, located on the +20' elevation of the Reactor Building -
- 13. NWCORRM-Northwest Corner Room, located on the -17'. elevation of the Reactor Building-
- 14. PIPETNL -
Pipe Tunnel, located between each Reactor Building and Turbine.
Building, Control ' Building, and Radwaste Butiding' on -3'-
elevation 15.RC Reactor containment 16 RHRARM Residual Heat Removal Pump-Room located on the -17' and
+20' elevations of the Reactor Building 88 1/39
Table 4-1.
Definition of Brunswick 2 Building and g
Location Codes (Continued)
Codes DescrIntions
- 17. RHRBRh1 Residual Heat Removal Puma Room located on the -17' and 420' elevations of the Reactor Building southeast corner-
- 18. RHRPIPECHASE RHR Pipe Chase the vertical and horizontal space connecting the Emergency Core Cooling System Pipe Tunnel and the two RIIR Pump Rooms on'the + 20 and + 5 elevations in each Reactor Building i
- 19. RWCUPRhi Reactor Water Cleanup Room, located on the +50i elevation of the Reactor Building
- 20. SThiTUN Steam Tunnel, located between each Reactor-Building and -
Turbine Building -
- 21. SWBLDG Service Water Building, located east of the Reactor Building and -
west of the intake Canal
- 22. SWCORRh1 Southwest Corner Room, located on the 17' elevation of the Reactor Buildirig -
23.
Turbine Building, located west of the Reactor Building
- 24. TOR Torus area, located on the 17' elevation of the Reactor Building:
4 4
4
- O t
j.
89-1/89 '-
l m.
- u. - _.._ _ _ _, _.._. :
I Table. 4 2.
Partial Liating' of Components by Location at '
Brunswick 2
)
j i
LOCA140N SYSTEM
.. COMP. 10
MOV j
1 20RB ECCS RHR 168 MOV 20RB ECC5 RNR-16B MOV 20RB ECCS RHR 21A MOV 20RB ECCS RHR 2tA -
MOV -
'i 20RB ECC6.
RNR 21B -
20RB EP EP MCC-1XDA MCC 20RB EP-EP MCC-1XDB MCC 20H6 EP EP MCC 2XA MCC 20RB EP EP MCC-2XA2 MCC 20RB EP EP MCC-2XB MCC-20RB EP EP MCC 2XB2 MCC 20RB EP EP MCC-2XC' MCC 20RB EP EP MCC-2XD MCC 20RB EP EP-MCC-2XDA MCC 20RB EP EP MCC 2XDB MCC 20RB RCIC RCIC-13 MOV t
23DGA EP EP DG-1 DG 230GA EP -
I.
230Gb EP.
- EP DG-3 DG-230GC.
- \\
23DGD EP EP DG-4 DG 23DGD EP EP MCC DGD MCC-23DGSG5 EP EP BS-ES BUS 23DGSGS EP EP TR F05 TRAN I
90 1/89'
- =
Table 4 2.
Partial Listing of Compoitents by Location at Brunswick 2 (Continued)
O
(
LOC A TION SYSTEM COMP, ID COMP.
TYPE 23DGSG6 EP EP-BS E6 BUS 23DGSGo EP EP TR FB6 TRAN 23DGbG7 EP EP BS E7 BUS 230GSUI EP EP TR-FBI TRAN 2JDG5Go EP EP-BS E8 BUS 23DGSGS EP E P-T R-F B2 TRAN 2CSA EP EP MCC-2CA MCC 2GSA EP EP MCC-2CB MCC 50DGbG1 EP EP BS-El BUS 50DGSG2 EP EP BS E2 BUS bODGSu3 EP EP BS E3 BUS bODG5G4 EP EP bS E4 BUS 50Rb ECCS CS 4A MOV 50RB ECCS CS 4B MOV 50kB ECCS CS 5A MOV v/
i 50RB ECCS-CS5B MOV BATIA EP EP BS 1A BUS BATIA EP EP BS 1A BUS BAT 1A EP EP BS-1 A BUS BATIA EP EP BS-1 A BUS BATIA EP EPBT1A BAT BATIB EP EP BS 1B BUS BAT 1B EP EP BS 18 BUS BAT 1B EP EP BS-2B BUS BATIB EP EP BSIB BUS E.118 EP EP GT 1B BAT BAT 2A EP EP BC-2A 1 BC BA12A EP EP bC 2A 2 BC BAT 2A EP EP BS 2A BUS BAT 2A EP EP BS 2A BUS
('N\\
()
BAT 2A EP EP BT 2A BAT-91 1/89
Table 4 2.
Partial Listing of Components by Location at 13runswick 2 (Continued)
O
/
4
'Q' LOCAllOf4 SYbTEM COMP. ID COMP.
TYPE BAi26 EP EP BC 261 BC BA128 EP EP BC 28-2 BC BAT 26 EP EP-BS-28 BUS BAl2B EP EP BS-2B BUS BAl2B EP EP Bi 20 BAT CSI ReiC CST TANK ECCSTNL ECCS HPC6 0 MOV ECCSTNL ECCS RHR-15A MOV ECCSTNL ECCS RHR 158 MOV ECCSTt4L RCiG RCIC4 MOV ECCSTNL RCS HPC6-3 MOV ECCSINL ROS RCIC-8 MOV ECCSINL RCS RMR-8 MOV HPCiRM ECCS HPCI 1 MOV HPCiRM ECCS HPCI-4 MOV
\\
HPCiRM ECCS HPC6-4 MOV HPCIAM GCCS HPCI-41 MOV HPCiRM ECCS HPCI42 MOV HPCIRM ECCS HPCb59 MOV HPC6AM ECCS HPCI-TP TDP HPCiHM ECCS HPCI-TSV8 HV MSIVRM ECCS RCS 32A MOV MStVRM RCIC RCS 32B MOV MStVRM RCS RCS 19 MOV f4WCORAM ECCS CSISA MOV P4WCORRM ECCS CS1A MOV NWCORHM ECCS CS P2A MOP RC ECCS HPCI-2 MOV RC RCIC RCIC-7 MOV RC RCS HPCI2 MOV p)
RC RCS RCIC 7 MOV t's RC RCS RCS-1 MOV 92 1/89
Tablo 4 2.
Partial Listing of Components by Location at Brunswick 2 (Continued)
)
k
/
~ LOC AllON SYSTEM COMP, ID COMP.
V TYPE RC RCS RCS16 MOV RC RCS RHR-9 MOV RnRARM ECCS HPC6 7 MOV kdMRM ECCS HPCI-4 MOV RmMRM ECCS RHR 10 MOV RNRARM ECCS RPR 10 MOV RHRARM ECCS NHR 20A MOV RHMHM ECCS RNA 3A MOV 7
RnRARM ECCS RHR 47A MOV RHRARM ECCS RH R-48A MOV Y
RHRARM ECCS RNA-48A MOV RHRARM ECCS RHR 4A MOV RMRARM ECCS RHR 40 MOV RMRARM ECCS RHR-6A MOV RMMRM ECCS RNR 6C MOV
(
RHMRM ECCb RNA HA2[
HX RHRARM ECCS RHR P2A MDP RMMAM ECCS RNR P2C MDP RHR3RM ECCS HPCI11 Mov THIBRM ECCS RHR 200 MOV RetRBRM ECCS RHA-38 MOV RHRBRM ECCS RHR 478 MOV RH ABRt
- ECCS RHR 40 MOV RHRBRM ECCS RHR4B MOV RHRORM ECCS RHR4D MOV RNRSRM ECCS RHR-HX2B HX RHRbHM ECCS RHR P2B MDP f~g RnksHM ECCS e
iR-P20 MDP k
)
N,/
. _ _ _ - _ - - - - - - - ~ - - - - - - - - -
Table 4 2.
Partial Listing of Componenth by Location at BrunsWlck 2 (Continued)
/
(
)
L OC A 140H SYSILM COMF, ID COMP.
l
(_/
RMRbRM R0lO RGiG-11 MOV i
Fwh6RM 6CIC RC;C12 MOV
~
RNRbRM NIC RCiG-19 MOV RNR6AM RCiG RC4C 22 MOV RHRBRM RCIC RCtC 29 MOV RHRBRM RGiG RGiG-31 MOV RnRBRM KiG RCiG-4$
MOV RMR64M RGiG RC4046 MOV RnhaRM RCiG RCIC TTVe MOV RMRPPECHASE ECCS RNR 17A MOV i
~
RHRP PECHASE ECCS RHR 178 MOV RWCUPRM RCS RCS-4 MOV SWB DG SWS CSW P2A MOP SWBLOG SWS CSW-P2B MDP g
t
(
)
SW6 LOG SWS CSW.P2C MDP -
%/
SWOLOG SWS NSW P2A MDP SWBLOG SWS NSW P2B MDP Sn0LDG SWS SWS 13 MOV SWBLDG SWS SWS 14 MOV SW6 LOG SWS SWS -15 MOV-SWOLDG SWS SWS 16 MOV SWBLDG SWS SWS-17 MOV SW6LDG SWS SWS 18 MOV SW6 LOG SWS SWS10 MOV SWBLOG SWS SWS-20 MOV SWCORRM ECCS CS-ISB MOV SWCORRM ECCS CS 1B MOV SWCORAM ECCS CS-P2B MDP TOR ECCS RHR 24A MOV TOR ECCS RHR 24A MOV
\\
/
%/
94 1/89
Table 4 2.
Partial Listing of Components by Location at Brunswick 2 -(Cont (nued).
LOCAllON SYSTEM COMP. ID COMP.
-TYPE
-i TOR
MO V.
?
.)
TOR.
ECCS RHR 28A MOV TOR ECCS RHR-26A MOV TOR-ECCS-RHR 288 MO V..
LOR ECCS RHR 288 MOV V255RM SWS SWS 255 MOV-I I
l 4
l1 l
l' i
l 3
1 i
i l
4 I
1
'.i i
l s
,\\ -
.)
1 95-1/89-
.i 1
Brunswick 1 & 2 5,
n[Ill,10cR Apjgy FOR llRUNSWICK
- 1. Doce, S.W., et al., "In. plant Soerce Terms Measurements at Bruswick Steam Electric Station," NUREG/CR.4245, EG&G Idaho, Inc., June 1985.
b h
I n
i j
i 96 1/89
"*e-"'"w-""rer-,.--.=
- tMg
1 i
Brunswick 1 & 2 APPENDIX A b
DEFINITION OF SYMilOLS USED IN TIIE SYSTEM AND LAYOUT DRAWINGS -
A1.
SYSTEM DRAWINGS -
5 t
A 1.1 Fluid System Drawings The simplifled system drawings are accurate representations of the major now paths
'in a system and the important interfaces with other fluid systems. As a general rule, small fluid lines that are not essential to the basic operation of the system are not shown in these drawings. Lines o_f this type include instrumentation-lines, vent lines, drain lines, and -
other lines that are less than 1/3 the diameter of the connecting major flow path. There usually are two versions of each fluip system drawingt a simplified system drawing, and a i
comparable drawing showing component locations. The drawing conventions used in the fluid system drawings are the following:
' Flow generally is left to right.
Water sources are located on the left and water " users" (i.e., heat loads) or discharge paths are located on the right.
One exception is the return flow path in closed loop sys: ems which is right tc, left.
Another exception is the Reactor Coolant System (RCS) drawing whlch-is
" vessel-centered", with the primary loops on both sides of the vessel.
Horizontal lines always dominate and break vertical lines.
Component symbols used in the Du!d system drawings are defined in Figure.
A-1.
Most valve and pump symbols are -designed to' allow the reader. to-distinguish among similar components based on their support system requirements (i.e., electric power for a motor or solenoid, steam to drive a turbine, pneumatic or hydraulic source for valve operation, etc.)
Valve symbols allow the reader to distinguish among valves that allow Dow -
in either direction, check (non return) _ valves, and valves that perform an-overpressure protection function. _ No attempt has been made to define the specine type of valve'(i e., as a globe, gate, butterfly, or other specific type -
of valve).
Pt.a.p symbols distinguish between centrifugal and positive displacement pumps and between types of pump drives (i.e.,nr, tor, turbine, or engine).
Locations are identined in terms of plant location codes defined in Section 4 of :
this Sourcebook.
Location is indicated by shaded " zones" that are not intended to represent'
-the actual room geometry.-
l Locations of discrete components represent the actual physical location ore l'
the component,-
Piping locations between discrete components represent-the plant areas 1
i through which the piping passes (i.e. including -pipe tunnels :and.
l underground pipe runs).-
Component locations that-are not known are indicated by placing the components in an unshaded (white) zone.
(
The primary flow path in the system is highlighted-(i.e., bold white line) in -
the location version of the fluid system drawings.
97.
- l/89 i
I
~ a
-,w.a,
. -~%
~... _.,, - -.. ~, _...,
......m
_ _ _ ~ - -. -
Brunswick 1 & 2 A 1.2 1:lectrical System Drewings
\\
The electric power system drawings focus on the Class IE port!ons of the plant's electric mwer system. Separate drawings are movided for the AC and DC portions of the Class lLi system. There often are two vers ons of each electrical system drawing: a simplified system drawing, and a comparable drawing showing component locations. The drawing conventions used in the electncal system drawings are the following-Flow per:erally is top to bottom in the AC power drawings, the interface with the switchyard and/or offsite
- rid is shown at the top of bl.? wing.
.n the DC powei hndags, the b. :eries and the interface with the AC power system are shown at the top of the drawing.
Verticallines dominate and break horizontallines.
Component symbols used in the electrical system drawings are defined in Figure A 2.
Locations are identified in terms of plant location codes defined in Section 4 of this Sourcebook.
Locations are indicated by shaded " zones" that are not intended to represca t'
aal room geometry.
- x. ions of discrete components represent the actual physicallocation of component.
't he electrical conn:ctions (i.e., cable runs) between discrete components, as shown on the electrical system drawings, DO NOT represent the actual
(_
cable routing in the plant.
Compoacnt locations that are not known are indicated by placing the
\\
discrete components in an unshaded (white) zone.
A2.
SITE AND LAYOUT DRAWINGS A2.1 Site Drawings A j;eneral view of each reactor site and vicinity is presentml alonj; with a simplified site plan slowing the arrangement of the major buildings, tanks, ud other features of the site. The general view of the reactor she is obtained from ORNL-NSiO 55 (Ref.1). The site drawings are ap?roximately to scale, but should not be used to estimate distances on the site. As built sea e drawings should be consulted for this purpose.
Labels printed in bold uppercase correspond to the locauon codes defined in Section 4 and used in the component data lirtings and system drawings in Section 3. Some additional labels are included for information and are printed in lowercase type.
A2.2 Layout Drawings Simplified building layout drawings are developed for the portions of the plant that cor....; components and systems that to described in Section-3 of this Sourcebook.
Generally, the following buildings are inewded: reactor building, auxiliary building, fuel building, diesel building, and the intake structure or pumphouse. Layout drawings genually are not developed for other buildings.
Symbols used in the simplified layout drawings are defined in Figure A 3. Major A
rooms, stairways, elevators, and doorways are shown in the simplified layout drawings I
however, many interior walls have been omitted for clarity. The building layout drawings, 98 1/89 l
{
Brunswick 1 & 2 y
are approximatel.v to scale should not be used to estims.te room size or distances. As built l
(
scale drawings for should.+e consulted his nurpose.
l Labels printed in upper, ce tolded also correspond to the location ccdes defined 'n l
j Section 4 and used in the component datt. listings and system drawings in Section 3. Some l
additional labels are included for information and are pnnted in lowercase type.
r A 3.
APPENDIX A REFERENCES i
1.
Heddleson, F.A., "D sign Data and Safety Features of Commercial Nuclear Power Plants.", ORNL NSIC 55, Volumes i to 4, Oak Ridge National Laboratorv, Nuclear Safety Infonnation Center, December 1973 (Vol.1),
l January 1972 IVol. 2), April 1974 (Vol. 3), and Mm.' '975 (Vol. 4)
N i
1 i
i a
I i
i 1
4
]
1 l
I I
I
)
)
'\\
99 1/89
-.n_.-----..
.~1,
..,.... _. -. _ ~.,..-..-.- --
..~.....
.~.
1
~
~
~
(OPEN CLC (D)
VALYL 3CV (OPEN CLO$tD)
O MOTOR *0PIR ATED VALyt. MOV h
NOTOR OPERAft0 N
(O P E N'C L O $ t D) k 3 W AY VALVE MOV (CLolt0 PORT MAY VARY) l t
V SOLEN 0ID OPER AttD VALVE. $0V SOLENOID.0PER ATED (CPEN8 CLOSED) 8 WAY V ALYC
- SOV (CLostD PORT MAY VARY) i HYDR AULIC VALVE e NV HYDAAULIO NON RETUAN (OPtHICLOSED)
V ALVE
- HCV (OPENrCL4?f D) p
,,,,,. PNtuMAtle V ALVE + NV PNEUMATIC NON RETURN r
(OPEN/Cl D$t D)
VALVE
- NCV (OPENrCLO$tD)
CHECK VALVE e CV N
S AFETy g ALyg, ty (CLO$tD)
O
Ch POWER OPER ATED RELitF VALVC, POWER OPER ATED RELIEF VALVE.
d 80LENotD PILOT TYPE. PORY d
PNEUMATICALLY OPER AftD a PORY (CLOSED)
OR-DUAL
- FUNCTION SAFETYtRELICF VALVE.SRV
_ (CLOSED) -
CtNTRIFUC AL CE NTRIFUC AL MOTOR DRIVEN PUMP e MDP TUR$1NE DRIVEN PUMP e TDP
\\
/
.l
.c I
POSITIVE DISPL ACE Mt HT MOTOR DRIVEN PUMP
- MDP POSITIVE DISPLACEMENT
=
TUR$1NT DRIVEN PUMP. TOP
.i i /
I i
l l
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Figure A 1. KeyTo Symbols in Fluid System Drawings 100 1/89_
...u..
.-a.
[
4
}
PWnewR MAIN CONDEN$tR COND RE AC10R Vt$$tt. RV I
L J
4 HEAT EXCHANotn. HX MECHANIC AL DR AFT COOLING TOWER h
h r--
STt AM.TO.W Af t R AIR C09 LINO UNif. ACU OR W ATER.TO.STE AM HE AT tsCH ANotR (i t. FttDW AttR 4
HE ATER, DR AIN COOLER, ETC.). HX n
fen l
tAa. Tx raaaaaaa SPRAY N0ZZLES. SN Y
v RUPfURE Di$K RD p
,p CRirl:t. OR Figure A-1, Key To Symbols in Fluid System Drawings (Continued) 101 1/89
f%
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l A C. 0:EStt otNtRe f CR 00
[
S ATTERY. D ATT
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OR A C. TURBINE GENE RATOR.10
=
l l
I C
on CIRCUlf PRf AKtR. CD E
g ),_qi og g),..g)
INTERL6CKt0 (OptN CLOSE D)
CIRCulf BRE AKERS CB
OR OR OTHER TYPE OF TR ANSFER SWITCH. ATS DISCONNECT DE VIC E CR (O P E N'CLO $ t 0)
MANUAL TRANSFER SWITCH. M1B SWITCHGt AR BUS
- BUS I W8 NAMI' }
MOTOR CONTROL CENTER e MCC OR N 7.D."
CR O!STniBVTION P ANEL. PNL I
l B ATTERY CH ARCCR (RECTiritR). DC 4
O I
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i M
en RELAY CONTACTS FU$t.FS (CPCN'CLOSCO) g ELEC1R,C,.0 TOR. M1e MO,0, 0,NCR A,0R. M0 (m\\
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Figure A 2.
Key To Sym'aols in Electrical System Drawings 102 1/89
. ~... -..
_ -. -.. - -. -. ~. - -...... -. -. - -. - - -.... - -.
i
)
i STAIRS SPIRAL T
$*,$P STAIRCASE own LADDER
(';
O = Up ELEVATOR D= Down
]
HATCH OR
.OPEN AREA GRATING DECK (NO FLOOR)
-O-DERSONNEL DOOR
- EQUIPMENT DOOR V
2 x
E RAILROAD TRACKS x
FENCE LINE
=
5 x
0 TANK / WATER AREA 1
+
g Figure A 3.
Key To Symbols in Facility Layout Drawings-103 1/89
Brunswick 1 & 2 APPENDIX 11 DEFINITION OF TERMS USED IN Tile DATA TAllLES Terms appearing in the dats @les in Sections 3 and 4 of this Sourcebook are defined as follows:
t SYSTEM (also LOAD SYSTEM) All components associated with a particular system 1
description in the Sourcebook have the same system code in the data base. System codes used in this Sourcebook are the following:
Cedr Definition' RCS Reactor Coolant System-RCIC Reactor Core Isolation Cooling System ECCS Emergency Core Cooling System (including HPCI, LPCI, CS and ADS)-
I&C Instrumentation and Cnntrol Systems EP Electric Powu System SWS Senice Water System COMPONENT ID (also LOAD COMPONENT ID) The component identification (ID) code in a data table matches the component 10 ihnt appears in the corresponding system drawing. The component ID generally begins with a system preface followed by a component number. The system preface is not necessarily the same as the system code described above. For component ids, the system preface corresponds to what the plant calls the component (e.g. IIPI, RHR). An example is HPI 730, denoting valve number 730 in the high pressure injection system, which is part of the ECCS. The component number is a contraction of the comsonent number appearing in the plant piping and instrumentation drawings (P&lDs) anc electrical one line system drawings.
LOCATION (also COMPONENT LOCATION and POWER SOURCE LOCATION) -
Refer to the location codes defined in Section 4.
COMPONENT TYPE (COMP TYPE) Refer to Table B 1 for a list of component type
- codes, POWER SOURCE - The component ID of the power source is listed in this field (see COMPONEN~l'ID, above), in this data base, a." power source"_ for a particular component (i.e. a load or a distribution component) is the next higher electrical distribution or-generating component in a distribution system. A single com aonent may have more than one power source (i.e. a DC bus powered from a battery and a battery charger).
POWER SOURCE VOLTAGE (also VOLTAGE)- The voltage "seen" by a load ofi power source is entered in this field. The downstream (output) voltage of a transformer, inverter, or battery charger is used.
EMERGENCY LOAD GROUP (EMERG LOAD GROUP) JAC and DC load groups (or electrical divisions) are defined as appropriate to the :)lant Generally, AC load grouos are identified as AC/A, AC/B, etc. The emergency loac group for a third of a-kind load (i.e. a " swing" load) that can be powered from either of two AC load groups would be -
identified as AC/AB. DC load group follows similar naming conventions.
[
104 1/89- -
d i
J
.,n,,.,
am,..
-,.w,,-
,.i.,,,--,,-.-,mn,
,,e,-
e.m-se-ec~ -, e
,- y w-e
-v
-,na-.,r
<w
-, -. - - *,-e n wwm e -tm U
I TAllLE 111.
COS1PONENT TYPE CODES COMPONFNT COMP TYPE VALVES:
hiotor operated valve hiOV Pneumati; (alt operated) valve NV or AOV liydraulie valve llV Solenoid operated valve SOY Manual valve XV Check valve CV Pneumatic non return valve NCV llydraulic non retum valve llCV Safety valve SV Dual function safety / relief valve SRV Power-operated relief valve PORV (pneumatie or solenoid operated)
PUMPS:
Motor-driven pump (centrifugal or PD) hiDP Turbine driven pump (centrifugal of PD)
TDP Diesel driven pump (centrifugal of PD)
DDP OTHER FLUID SYSTEM COMPONENTS:
Reactor vessel RV f
Steam generator (U tube or once through)
SG lleat exchanger (water to water HX, I-IX or water to-air liX)
Cooling tower CT Tank TANK or TK Sump SUMP Rupture disk RD Orince ORIF Filter or strainer FLT Spray nozzle SN Heaters (i.e. pressurizer heaters)
HTR VENTILATION SYSTEM COMPONENTS:
Fan (motor driven, any type)
FAN Air cooling unit (air to-water HX, usually ACU or FCU including a fan)
Condensing (air conditioning) unit COND EMERGENCY POWER SOURCES:
Diesel generator DG Gas turbine genert. tor GT Battay BA'IT C
r
\\
105 1/89
f TAllLE 11 1.
C051PONENT TYPE CODES (Continued)
N C051PO NENT CO31PTYPE ELiiCTRIC POWER DISTRIBUTION EQUIPMENT:
Bus or switchgear DUS Motor control center MCC Distribuuon panel or cabinet PNL or CAD Transformer TRAN or XFMR Battery charger (rectifier)
BC or RECT invener INV Uninterruptible power supply (a unit that may UPS include battery, battery charger, and inverter)
Motor generator MG Circuit breater CB Switch SW Automatic transfer switch ATS Manual transfer switch MTS
(
I OV 106 jfg9
.