ML20059B834
| ML20059B834 | |
| Person / Time | |
|---|---|
| Site: | 05200004 |
| Issue date: | 02/08/1991 |
| From: | Fleming W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20059B690 | List: |
| References | |
| 23A6723, NUDOCS 9310290112 | |
| Download: ML20059B834 (77) | |
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GENuclewEnergy ygasng aev A Revision Status Sheet Document Title Composite LC9Cnd_0LD11CIiption of Gropos Type Desien Specification FMF SBWR MPLitem No All-5299 Revisions C
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- 1. CONTENTS AND CONFIGURATION.
...............7 1.1 Scope and Purpose.
... 7 1.2 Document Strutture..
.7 1.3 Applicab!c bocuments...........
..8 l.3.1 Supporting Documents..
..8 1.3.2 Supplemental Documents.....
....... 9 1.3.3 Appended Data Sheets..
.. 9 1.3.4 laws and Regulations.
....9 1.3.5 Codes and Standards...
.9 1.4 Acronym s........
.9 1.5 Definitions..
. 14
- 2. GENERAL REQUIREMENTS AND CRITERIA...
l
.15 2.1 Ocneral Design Requirements......
.15 2.1.1 Units of Measure for Design.....,
............. 15 2.1.1.1 Primary System of Units..
.... 15 2.1.1.2 Secondary System of Units....
.. 20 2.1.1.3 Conversion Factors..
.. 20 2.1.2 Documentation.
2.2 Powcr Generation Design................
.. 23
. 24 2.2.4 1x>ad Following Capability..
. 24 2.3 Safety Design......
... 25 2.4 Severe Accident Mitigation.
. 33 2.5 Station Blackout Capability.
............ 33 2.6 ATWS Pr ovisions....
.... 34 2.7 Man. Machine Interfaces (MMI)...
.34 2.8 Plant Configuration and Security..
..... 35 2.9 Site Environmental Characteristics...............
.... 37 210 Design, Construction and Transportability..
......... 37 2.10.1 Marcrials and Processes.
... 37 2.10.2 Mechanical and Structural Loading.
....... 3 7 2.10.3 Mcchanical Design Requirernents.....
.40 2.10.4 Electrical Design Rcquirements...
... A0 2.10.5 Instrumentation and Control Design Requirements...
.40 2.10.6 Transp
.o
" Req uirements..................
. 40 2.11 Plant Duty Cy,
.d Life.....
... 40 2.12 Sy:. tem Interft.
. 41 2.13 Tc3tability of S r v and Cornponents.
.. 41 2.14 Quality Assura.
. 41 2.15 Reliability, Availaaility and Maintainability..
.. 44 ao w msv m
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Table oiCon tents (Continued)
Page 2.16 Control Room Habitability..................................
.. 44 2.17 Classification of Structures, Components, Parts, and Systems........... 45
.45 2.17.1 Safety Classification....
2.17.2 Quality Group Classification........
....... 45 2.17.3 Seismic Classification...
.. 45
.46 2.17.4 Electrical Classification...
2.17.5 Classification Related to Quality Assurance....
.... 46
- 3. REQUIREMENTS FOR STSTEMS................
...... 4 7
..47 3.1 Nucicar Steam Supply Systems.........
3.1.1 Reactor Pressure VessJ (BI1)...............
..... 47 3.1.2 N uclear Boiler (B21 )...........................
.........47
.48 3.1.3 Isolation Condenser (B32)...
3.2 Control and Instrumentation Systems...
. 49 3.2.1 Rod Control and Information (C11)......
.. 49 3.2.2 Control Rod Driw (Cl 2)...............
. 49 3.2.3 Leak Detection and Isolation (C21)...
.... 50 3.2.4 Feedwater Control (CSI).........
... 51 3.2.5 Standby Liquid Control (C41)..
. 51
.. 51 3.2.6 Neutron Monitoring (C51)..
3.2.7 Remote Shutdown (C61)...
............. 5 2 3.2.8 Multiplexing System (C62)..
... 52 3.2.9 Reactor Protection (C71,................
.. 52 3.2.10 Safety System Logic & Control (C74).........
. 52 3.2.11 Automatic Power Regulator (CS2)........
.. 53 3.2.12 Steam Byp.ss and Pressure Control (C85).
.. 54
...... 54 3.2.13 Process Computer (C91)...
............ 5 5 3.3 Radiation Monitoring Systems..
3.3.1 Process Radiation Monitoring (D11)..
. 55 3.3.2 Area Radi
',n Monitoring (D21).
... 55 3.3.3 Containn
- *mospheric Monitoring (D23)............
....... 55
... 55 3.4 Core Cooling Sy>
o 3.4.1 Gravity-Dnvc. woling (E50).
. 55 3.5 Reactor Servicing Equipment (F-series).
. 56 3.6 Decay IIcat Removal Network..............
............... 5 7 3.6.1 Fuel and Auxiliary Pools Cooling (G21)....
. 57 3.6.2 Reactor Water Cleanup / Shutdown Cooling System (G31)..
.. 57 i
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23A6723 swNo.
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Table of Cnr tents (Continued)
Page 3.7 Control Panc13..
... 58 3.7.1 Main Control Panels (H10)...................................................58 3.7.2 Main Control Room (MCR) Equipment Room Panels (H11)..... 58 3.7.3 Radioactive Waste (Radwaste) Control Room Panels (H14)......... 58 3.7.4 Local Panels and Racks (H21).
. 58
)
3.8 Nuclear Fuel (j-series).
.59 3.9 Radioactive Waste Management Systems......................
.................60-3.9.1 Liquid and Solid Radwaste (K10, K20).....
... 60 3.9.2 O ffgas ( R30 )........................................
...........................60 1
3.10 Power Cycle Systems.
. 61 3.10.1 Condensate and Feedwater (N21)...
. 61
? 10.2 Heater Drain and Vent (N22)........
......................62 3.10.3 Condensate Purification (N25)...............
............ 62 3.10.4 Main Turbine and Auxiliaries (N31 to N39, N11)............... 63 3.10.5 Generator and Auxiliaries (N41 to N51).
......M 3.10.6 Main Condenser and Auxiliaries (N61).
..M 3.10.7 Circulating Water (N71).
. 65 3.11 Station Auxiliary Systems....
. 65 3.11.1 Makeup Water (P10)..
. 65 3.11.2 Reactor Component Cooling Wr.ter (P21)........................... 66 3.11.3 Turbine Building Cooling Water (P22).
...f4 3.11.4 Chilled Water System (P25)..........
. 66 3.11.5 Cemdensate Storage and Transfer (P30) _..
..f6 3.11.6 Oxygen Injection (PS2)......
....f6 3.11.7 Process Sampling (P33)..........
.,66 3.11.8 Reactor Semcc Water (P41).......
... 66 3.11.9 Turbine Building Senice Water (P42).
. 66 3.11.10 Service Air and Instrument Air (P51, P52).
'....... f6 3.11.11 High Pressure Nitrogen Gas Supply (P54)...
. 66 3.11.12 Auxiliary Boiler (P62).
..f4 3.11.13 Hydrogen Water Chemistry (P73).
... 67 3.11.14 Post Accident Sampling (P91).
. 67 3.11.15 Iron Injection (P95)......................
..................67 3.12 Station Electrical Systems...............
. 67 3.12.1 Electrical Power Distribution (R10).............
. 67 3.12.2 Unit Auxiliary Transformers (Rll)..
.. 67 3.12.3 Isolated Phase Bus (R13).......
. 67 3.12.4 Non-Segregated Phase Bus (R21)..
.. 67 3.12.5 Metal Cbd Switchgear (R22)........
.... 67 3.12.6 Power Cen ters (R23)...............................
. 68 7
3.12.7 Motor Con trol Centers (R24)...................
...............68 NF 0 acNRE v 448) l
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23A6723 s u o.
5 AEV A Table of Contents (Continued) i faEs 3.12.8 Raceway (R31)...
. 69 3.12.9 Cable (RSS).......
.70 3.12.10 Plant Grounding (R34).........
................................70 3.12.11 Electric Penetrations (R35)....
. 71 3.12.13 Standby AC Power Supply (R40)......
........ 71 3.12.14 Direct Current Power Supply (R42)....
........... 72 3.12.15 Vital (Uninterruptible) AC Power Supply (R 16)...
. 72 3.12.16 Instrument and Control Power Supply (R47).......
. 72 3.12.17 Communication (R51)....
.73 3.12.18 Lighting and Servicing Powcr Supply (R52).
. 73 3.13 Power Transmission Systems (Sseries).
.. 73 3.14 Containment and Environmental Control Systems..
.. 74
. 74 l
3.14.1 Containment System (T10)...
3.14.2 Containment Vessel and Structures (Tl1, T12).
. 74
.75 3.14.3 Passive Containment Cooling Sptem (T15).
3.14.4 Atmospheric Control (T31)........
. 75 3.14.5 Drywell Cooling (T41).....
....... 75 3.14.6 Flammability Control (T49).....................
. 75 3.14.7 Suppression Pool Temperature Monitoring (T53)............... 76 3.15 Structures and Servicing Systems......
. 76 3.15.1 Turbine Pedestal (U24).........
.76 3.15.2 Crancs, Hoists and Elevators (U31)...
. 76
.76 3.15.3 Turbine Building 11VAC (U39).
3.15.4 Reactor Building IIVAC (U40)..
. 76 3.15.5 Other Building IIVAC (U11).....
.. 77 3.15.6 Fire Protection (U43)....................
... 77, 3.15.7 Buildings & Services (U65 to U73, U42, U44, U45, U50)...
.. 78 3.16 Intake Structure and Servicing Equipment (W. series).
.. 78 3.17 Yard Structures and Equipment (Yu ries).
.78 AFO*'.iaM ve'ut
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2SA6723 sa No.
6 acv A i
l List of Tablej i
Page Table LSI Nuclear Regulatory Commission (NRC) Code of Federal Regulations (CRF)..
-.10 Table 1.3-2 SBWR Design Documentation Standards.
10 i
Table 1.4-1 A cron ym s............................
.....I1 Table 2.1.1 -1 SI Base Un i ts..........................................
............ 15 Table 2.1.1-2 SI Supplementary Units...
.16 Table 2.1.1-3 SI Derived Units with Special Names..
.16 Table 2.1.1-1 Other SI Derived Units......................
.16 Table 2.1.1-5 Decirnal Fraction and Muldelus of SI Units Having Special N a m e s...
18 Table 2.1.1-6 Degree Celsius.......
..18 Table 2.1.17 Acceleration of Gravity....
.18 Table 2.1.1-8 Radiation Dose, Neutron Flux and Fluence...........................
.19 Table 2.1.1-9 Prefixes for SI Units......
.19 Table 2.1.1-10 Examples of Units in U. S. Customary System......
...20 i
Table 2.1.1-11 Radiation Dose Neutron Flux and Fluence....
................... 20 Table 2.1.1-12 Conversion Factors....
21 Table 2.3-1 Simultancous Design Basis Events...........
.........,... 28 Table 2.S2 Functional Classi0 cation of Systems Design Ilasis........
..................30 Table 2.9-1 Envelope of Plant Site Environmental Characteristics......................38 Table 2.121 Protective and Engineered Safeguards..
42 Table 2.17-1 Classification of Structures, Components, Parts,;.d Systems..
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- 1. CONTENTS AND CONFIGURATION 1.1 Scone and Pumose 1.1.1 The Composite Design Specification establishes top-tier requirements and criteria for design of the Simplified Boiling Water Reactor (SDWR) Standard Plant. The SBWR Standard Plant includes all buildings and structures (except for the main transformer, switch yard, cooling tower, and water treatment building, and demineralized water storage tanks), along with associated housed equipment and access controls, for the overall power block shown in the plot plan of the accompanying appended data sheet (para.1.3.3.a)to this specification.
1.1.2 The Composite Design Specification provides requirements which form the overall design basis of the plant. It also defines requirements where appropriate for major groups of systems and for individual systems where intuactive functions are significant in the Nuclear Steam Supply System (NSSS) and the Balance of Plant (BOP). The~ systems structure diagram referenced in paragraph 1.3.1.b shows the systems groups and individual systems which are described in this specification.
1.2 Document Structurg 1.2.1 The Composite Design SpeciDcation contains principal design requirements, criteria and guidelines. It receives project definition inputs. It is supported and supplemented by the d6coments listed in paragraphs 1.3,1,1.3.2, and 1.3.3, respectively. It applies licensing requirements, topical reports and industrial codes and standards by reference as applicable (paragraphs 1.3.4 and 1.3.5). It provides performance and functional requirement outputs which guide,in turn, the development of System Design Specifications (SDSs) for the individual systems comprising the SBWR Standard Plant, 1.2.2 The Composite Design Specification consists of a base specincation (this document) plus a group of appended data sheets (para 1.3.3) pioviding important performance data and system functional req >2irements for reference as well as design control and applicable Standard Review Plans. Regulatory GJ tes, and codes and standards.
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8 ntv A 1.3 Applicable Documents 1.3.1 Eupportine Documents. Supporting documents are those documents that complete the requirements of this specification and are refered to herein.
Designation -
a.
SHWR Standard Plant Master Parts List 18NS07A04 b.
SilWR Systems Structure A10-1010 Plant General Requirements System Application A11-1010 c.
d.
Flood Protection Requirements A112002 -
Wind & Tornado Protection Requirements All-2003 c.
f.
Missile Protection Requirements All-2004 g.
I.ightning Protection Requirements All-2008
_l h.
Fire Protection Requirements All-2009 i.
Pipe Rupture Protection Requirements A11-201S j.
Jet Impingement Protection Requirements A11-2014 k.
Mechanical Equipment Separation A11-2018 1.
Electrical Equipment Separation A11-2019 Equipment Environmental Data A11-2020 m.
load Combinations and Acceptance Criteria All-2032 n.
o.
Materials and Process Control All-2043 p
Steady State Performance Requirements All-3005 q.
Transient Performance Requirements All-3006 ECCS Requirements All-3007 r.
Reactivity Control Requirements A11-3008 s.
3 1.
Plant Duty Design Requirements All-3009 Water Quality A11-4000 u.
Plant System Interfaces All-1011 v.
w.
Reactor licat 11alance A11-5282 GETAn Operating Limits All-5285 x
y.
Turbine Ileat 11alance All-5300 Reliability, Availability and Maintainability (RAM) Criteria and A18-1020 z.
Guidelines Design of Sptem Controls and Instrumentation A32-1030 aa.
ab.
Generic Operation & Maintenance Requirements A80-8010 ac.
SilWR Standard Safety Analysis Report A90-1040 ad.
SI1WR Design and Certification Program Quality Assurance Plan NEDG 31831 j
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23A6723 s No.
9 ntv A 1.3.2 Sunnlemental Documents. Supplemental documents are those documents that are to be used in conjuction with this specification but are not needed to complete the requirements specified herein. None.
1.3.3 Appended Data Sheets. The following datasheets form a part of the requirements of this doc ument.
Desicnation Composite Design Specification - Design and Configuration Data All-5299 2.
Sheet 23A6723AA h
Composite Design Specification - Standard Review Plans and All-5299 Regulatory Guides Data Sheet,23A6723All c.
Composite Design Specification - Codes and Standards Data Sheet, A11-5299 23A6723AC 1.3.4 I.aws and Reculations Nuc1 car Regulatory Commission (NRC), Title 10, Code of Federal Regulations (CFR) -
a.
see Table 1.3-1 for those referenced to herein.
b.
Nuclear Regulatory Commission (NRC) Standard Review Plans (SRPs) and Regulatory Guides (RGs)- as required by the document referenced in paragraph 1.3.3.b.
SI1WR Licensing Review Information Document A90-1030 c.
1.3.5 Codes and Standard,3 Industrial Codes and Standards-as rcquired by the document referenced in a.
paragraph 1.3.3.c.
b.
Si1WR Design Documentation Standards - see Table 1.3-2.
1.4 Acronyms See Table 1.4-1 e.to acnm v m, p;-
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i Table 1.31 Nuclear Reculatory Commission (NRC) Code of Federal Reculations (CRF)
Nom Regulations refered to in this specification, not a complete list of Regulations 10CFR20, Standards for Protection Against Radiation a.
ti 10CFR21, Reporting of Defects and Noncompliance 10CFR50, Domestic Licensing of Production and Utilization Facilitics c.
d.
10CFR71, Packaging and Transportation of Radioactive Material 10CFR73, Physical Protection of Plants and Materials e.
f.
10CFR100, Reactor Site Criteria 1
-Table 1.52 SilWR Desien Documentation Standards a.
A00-3010 Piping and Instrumentation Diagram Standard 1
Piping and Instrumentation Diagram Symbols 1
b.
A00-3020 Process Flow Diagram Standard c.
A00-3030 Instrument and Electrical Diagram Standard d.
A00-3040 Logic Diagram Standard
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c.
A00-3050 System Design Specification Standard f.
A00-3060 Equipment Requirements Spe'cification Standard g.
A00-3070 Equipment Drawings (Assembly and Outline) Standard h.
A10-1050 CRT Display Conventions NTO 947 PE/ CRA!
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Table 1,4-1 Acronyms ABS Auxiliary Boiler System AC Alternating Current ACS Atmospheric Control System ALWR Advanced Light Water Reactor APRS Automatic Power Regulator System i
ARI Alternate Rod Insertion l
ARMS Area Radiation Monitoring System ASD Adjustable Speed Drive ASME American Society of Mechanical Engineers ATWS Anticipated Transients Without Scram BOP Balance of Plant BTP Branch Technical Position BWR Boiling Water Reactor C&FS Condensate and Feedwater Sptem C&I Control and Instrumentadon CAMS Containment Atmospheric Monitoring System CFR Code of Federal Regulations CIRC Circulating Water System CPS Condensate Purification System CRD Ccmtrol Rod Drive CS&TS Condensate Storage and Transfer System CV C<mtainment Vessel CWS Chilled Water System
-i DC Direct Current l
DCS Drywell Cooling System DPV Depressurization Valve EPDS Electric Power Distribution System EPRI Electric Power Research Institute FAPCS Fuel and Auxiliary Pools Cooling System FCS Flammability Control System FMCRD Fine Motion Control Rod Drive FPS Fire Protection System FTS Fuel Transfer System FWCS Feedwater Control System GC3 Generator Cooling System GDCS Gravity-Driven Cooling System GETAB General Electric Thermal Analysis Basis GSOS Generator Sealing Oil Sptem
!!DVS IIcater Drain and Vent System HGCS I-lydrogen Cas Cooling System M 0 It07 IHL v Cnt)
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IIPNSS High Pressure Nitrogen Supply System IIVAC IIcating, Vendlation and Cooling HWC Hydrogen Water Chemistry System IAS Instrument Air System IASCC Irradiation Assisted Stress Corrosion Cracking IBD Interlock Block Diagram IC isolation Condenser ICS Isolation Condenser System IGSCC Intergranular Stress Corrosion Cracking US Iron Injection System LD&lS Leak Detection and Isolation System LOCA Ims of Coolant Accident IDOP Loss of Off-Site Power LPRM Local Power Range Monitor LWS Liquid Waste System MCPR Minimum Critical Power Ratio MCR Main Control Room MLHGR Maximum Linear Heat Generation Rate MMI Man-Machine Interface MPL Master Parts List MSiv Main Steam isolation Valve MWS Makeup Water System NBR Nuclear Boiler Rated NBS Nuclear Boiler System NMS Neutron Monitoring System NRC Nuclear Regulatory Conunissinn NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake OG Offgas System O!S Oxygan Injection System P&lD Piping and Instrumentation Diagram PAM Post Accident Monitoring PASS Post Accident Sampling System PCCS Passive Containment Cooling System i
PRA Probabilistic Risk Assessment PRMS Process Radiation Monitoring System PRN M Power Range Neutron Monitoring PSS Process Sampling System RAM Reliability, Availability and Maintainability RCalS Rod Control and Information System RCCV Reinforced Concrete Containment Vessel A.IO 807 FEV CRE}
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23A6723 SH NO.
13 atv A Table I A-1 Acronyms (Continued)
RCCWS Reactor Component Cooling Water System RCPB Reactor Coolant Primary Boundary RG Regulatory Guide RPS Reactor Protection System RPV Reactor Pressure Vessel RSS Rernote Shutdown System RSWS Reactor Service Water System RWCU/SDC Reactor Water Cleanup / Shutdown Cooling SAS Senice Air System SB&PC Steam Bypass and Pressure Control System SBWR Simplified 13 oiling Water Reactor SDC Shutdown Cooling SDS System Design Specification SLCS Standby Liquid Control System SPTMS Suppression Pool Temperaturc Monitoring System SRNM Startup Range Neutron Monitor SRO Senior Reactor Operator SRP Standard Review Plan SRV Safety Relief Valve SSE Safe Shutdown Earthquake SSIL Safety System Logic and Control SWS Solid Waste System TASS Turbine Auxiliary Steam System TBCWS Turbine Building Cooling Water System TBS Turbine Bypass System TBSWS Turbine Building Senice Water System TCS Turbine Control System TGSS Turbine Gland Seal System TIDS Turbine Lubricating Oil System TMSS Turbine Main Steam System TSC Technical Support Center VAC Volts Alternating Current VDC Volts Direct Current NEO 6C7 iREV 4.Bes
GEhdwEmpy 234c223 soo-i4 nrv A 1.5 Definitions
].5.1 Safety-Related. A plant structure, system, or component, or part thereofis safety-related ifit is necessary to assure:
The integrity of the reactor coolant pressure boundary; a.
b.
The capability to shut down the reactor and maintain it in a safe shutdown condition; or The capability to prevent or mitigate the consequences of accidents which could result in c.
potential ofr-site exposures comparable to those referred to in 10CFR100 (Ref. Table 1.S1).
Safety-related is synonymous with the term "Ilasic Component" used in 10CFR21 (Ref. Table !.SI).
1.5.2 Non-Safety-Related. A non-safety-related plant structure, system or component, or part thereof,is that which does not meet any of the definitions for safety-related (Ref. para.1.5.1).
1.5.3 Special Capability. A non-safety-related plant structure, system or component, or part thereof that are subject to supplemental regulatory or QA requirements is cla.ssified as special capability.
1.5.4 Severe Accident. A severe accident is an event or sequence of events beyond the licensing basis in which substantial damage is done to the reactor core whether or not there are severe oft-site consequences.
1.5.5 Active Comnonent. An active component is a component in which mechanical movement must occur to accomplish the function of the component.
1.5.6 61tive Faibits. An active failure is a malfunction, excluding passive failures, of a component that relies on mechanical movement to complete its intended function upon demand.
i 1.5.7 Panive Component. A passive componentis a component that is not an active j
component.
1.5.8 Passive Failure. A passive failure is the blockage of a process flow path or failure of a component to maintain its structural integrity or stability, such that it connot provide its intended function tmon demand.
1.5.9 Single Failure. Single failure is an occurrence of either:
A failure of any active component (assuming passive components function properly), or a.
b.
A single failure during long-term conditions of any passive component (assuming active components function properly).
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23A6723 ssNo. 15 nsv A
- 2. (TENERAI. REQUIREMENTS AND CRITERIA 2.1 D_q_rteral Design Requirements 2.1.1 Units of Measure for Desien. The units of measure used in all documents related to the SBWR shall be the primary system of units defined below. A secondary system of units, also denned below, will be provided in parentheses after the primary units for all significant dimensions and other parameters in order to promote understanding by all parties. In the case of reports and other documents which contain detailed calculations, only important results and conclusions are required to have the values in the secondary system of units shown along with the primary system. An exception to the use of the primary and secondary systems of units described above will be in the case of a document which is based on prescribed analpes (such as the ASME Code) or computer codes which are defined in the secondary system of units. In this case the secondary system may be used throughout with the important results and conclusions shown with the primary system of units in parentheses.
If a third system of units is used to generate such documents then both the primary and secondary systems of units will also be shown for the important results and conclusions.
2.1.1.1 Primary System of Units. The primary system of units shall be the International System of Units, as adopted by the Conf 6rence G6n6 rale des Poids et Mesures, or St. The SI base units are the meter, kilogram, second, ampere, kelvin, candela, and mole.
Supplementary and derived units in SI may also be used along with others defined below.
The reference for the use of the Si system is " Standard Practice for Use of the International System of Units", American Society for Testing and Materials (ASTM) E380. The SI system of units is defined in Tables 2.1.1 -1 throt.gh 2.1.1-9 Tabic 2.1.1-1 Si Base Units Physical quantity Name of SI Unit Symbol for SI Unit Length meter m
Mass kilogram kg Tim e second s
Electric current ampere A
Thermodynamic temperature kelvin K
Amount of substance mole mot Luminous intensity candela cd NE O 807 $rv 49) 4
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nsv A Tabic 2.1.1-2 SI Suppkmentary Unita Physical cuantity Name of SI Unit Syrnhol for SI Unit Plane angle radian rad Solid angic stcradian sr Table 2,1.13 S1 Derived Units with Special Names Name of Symbol Phvsical quantity SI Unit for SI Unit De6nition of SI Unit Force newton N
m kg s-2 Pressure, stress pascal Pa m 1 kg s-2 (=N m-2) m kg s 2 (=N m) t Energy joule j
m.kg s 3 (-J s-1) 2 Power watt W
Electric charge coulomb C
sA m kg s 3 A-1 (=J A-1 s-1) 2 Electric potential difTerence volt V
m kg s-3 A-2 (,y.g-1) 2 Electric resistance ohm D
Electric conductance siemens S
m-2 kg'l s A2 (,A.y 1 g-1) 5 Electric capacitance farad F
m-2 kg-1 s A2 (=A s V-1) 4 m, y.3-2.gl (.y.3) 2 Magnetic flux weber Wb g
m y
,3 2 A-2 (=V A 1 s) 2 Inductance henry H
g 2
1 2
Magnetic flux density te sla T
kg r N (=V s nr )
1,uminous flux lumen im cd sr Illuminance lux lx m-2 cd sr Frequency hertz liz s-1 Activity (of radioactive source) becquerel Ik; s-1 Absorbed dose (of radiation) gray Gy m r2 (,j,yg 1) 2 Table 2.1.1-4 Other Si Derived Units (not an exhaustive list)
Symbol for Physical quantity SI Unit SI Unit 4
Absorbed dose rate gray per second Gy s-1 Acceleration meter per second squared m s-2 Angular acceleration radian per second squared rad r' Angular velocity radian per second rad s 1 2
Area square meter m
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17 ntv A Tabic 2.1.14 Other Si Derived Units (Continuedl Symbol for Physical quantity SI Unit SI Umt Concentration mole per cubic meter mol m-3 Current density ampere per square meter A. m 2 l
Density kilogram per cubic meter kg m 8 Dynamic viscosity newton-second per square meter Nsm2 Electric charge density coulomb per cubic meter Cm3 Electric field strength volt per meter Vm!
Electric flux density coulomb per square meter C m2 Energy density joule per cubic meter Jm3 Entropy joule per kehin J K3 11:at Capacity joule per kehin J.K'l 11ea flux density, irradiance watt per square meter W m"2 m. 3-t r
Kinematic viscosity, diffusion square meter per second coeflicient Luminance candela per square meter cd m-2 Magnetic field str'ength ampere per meter Am1 Magnetic permeability henry per meter H tn-1 Molar energy joule per mole J mol 2 Molar entropy, molar heat joule per kehin mole J.K 1 mol 1 capacity Permittivity farad per meter Fm1 i
Power density watt per square meter Wm2 Radiance watt per sq. meter steradian W m r sr1 Radiant intensity watt per steradian W sr-1 Specinc energy, enthalpy joule per kilogram J kg 1 Specific entropy joule per kilogram kehin J kg1 K1 Specific heat capacity joule per kilogram kehin J kg-1 K 1 m kg !
8 Specinc volume cubic meter per kilogram Thermal conductivity watt per meter kehin W m1 K1 Torque, moment newton meter Nm Velocity meter per second ms3 3
Volume cubic meter m
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lo to. m 1& 8 me 2
Cross section barn b
Volume liter 1L 10-8 m8 Mass tonne t
103 kg 5
Pressure bar bar 10 Pa Tim e minute min 60 s Time hour h
3600 s Time day d
86400 s Angic degree (n/180) rad Angle minute (n/10800) rad Angie second (n/648000) rad Table 2.1.1-6 Degree Celsius Degree celsius is an acceptable substitute for kehin ir. cases of relative temperature (not absolute temperature).
Physical <piantity Name of Unit Symbol for Unit Definition of Unit Celsius temperature degree Celsius
'C
- C=K Tabic 21.17 Arceleration of Gravity Gravity varies over the surface of the earth, F.owever, a standard measure of the acceleration of gravity has been adopted called, g. This measure should only be used in ficids where it is customarily used to denote units of acceleration relative to standard acceleration of gravity such as seismology.
Physical cuantity Symbol for Unit Definition of Unit Free fall acceleration, stand?rd g
9.806 650 m.s'2 NE O B07 (REV d?SS) f s m 4
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Table 2.1.1-8 Radiation Dose, Neutron Flux and Fluence The prirnary unit of radiation dose shall be the sievert which is defined as:
Radiation dose in sieverts = RBE x Absorbed dose in gray RBE is the Relative Biological Effectiveness, a factor which varies depending on the type and nature of the absorbed radiation Physical quantity Narne of Unit Symbol for Unit Definition of Unit Radiation dose sicvert Sv RBE x gray (m.32 orJ kg-1) 2
'Ihe following shall be the only units of measure for neutron flux and fluence:
Physical quantity Name of Unit Symbol for Unit Neutron flux neutrons per sq. cendmeter second n c m 2 3-1 Neutron fluence neutrons per sq. centimeter n cm 2 Table 2.1.1-9 Prefixes for SI Units Muhi lication l' actor Prefix Symbol P
1038 exa E
1015 peta P
1 10 r tera T
9 10 giga G'
6 10 mega M
10' kilo k
102 hecto h
103
-deka da 1RI deci d
lgt centi c
4 10 milli m
1&*
micro 9
10 nano n
tal pico p
1035 femto f
1&l8 atto a
nro m inrv em 1
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Customary units as defined in Tables 2.1.110 and 2.1.1 11 Table 2.1.1-10 Examples of Units in U. S. Custona.ryJystem Physical cuantity Name of Unit Symbol for Unit 1.cngth inch, foot, or mile in, ft, or mi Mass pound mass ihm Force, weight pound force or ton (short) lbf or ton = 2000 lbf
- Tim e second s
lleat or energy British Thermal Unit or foot pound Btu or ft lb -
Power horsepower or foot pound per second hp or ft lbs sec'l Teinperat ure Fahrenheit
'F Absolute temperature Rankine R
Volu m e cubic feet or gallons ft8 or gal Torque, moment foot pounds ftlbs If confusion can exist between the unit ton in the U. S. Customary system and the unit tonne in the SI system then the phrase "short" shall be appended to the unit ton,i.e. ton (short) or short ton.
Table 21,1-11 Radiation Dose. Neotion Flux anri Finenre The secondary unit of radiation dose shall be the Rem which is defined as:
Radiation dose in Rem = RBE x Absorbed dose in Rad 1
Rad =.01 Jou!c.kg of absorbed dose The units of measure for neutron flux and fluence shall be the same as defined for the primary system.
2.1.1.3 Conversion facints. The factors for converting between the primary system of units and the secondary system of units shown on Table 2.1.112 shall be consistently used. For conversion factors not covered by the tables below and further guidance on their application the " Standard Practice for Use of the International System of Units", ASTM E380, shall he used as a guide.
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Table 2.1.1-12 Conversion Factors To convert from secondary SulfM 10 primary System Multiply by Acceleration in s.2 m s*2 2.540 000 x 1gr ft rr a n. s.2 3.048 000 x 10-1 h
int r
6.451600 x 104 m
ft!
r 9.290 301 x 102 m
Bendinc moment or torouc Ibfin N.m 1.129 848 x 1&1 0
lbf ft Nm 1.355 818 x 10 EngIrv or work Btu J
1.055 056 x 105 calorie J
4.186 800x 100 ft lbf J
1.355 818 x 100 Specific encrev. enthalov j
Utu lbm'1 J. k g-1 2.326 000 x 108 EDItc lbf N
4.448 222 x 100 ton (shon)
N 8.896 444 x 108 kip (1000 lbf)
N 4.448 222 x 105 1.cn c th ft m
3.048 000 x 101 in m
2.540 000 x 1gr mile (U. S. statute) m 1.609 347 x 108 Mass 1bm kg 4.535 924 x 101 f i slug or Ibf s fr kg 1.459 390 x 101 j
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To convert from secon dary_. system 19_ primary systen1 Multinly by Mass oct Unit Volume i
lbm fr5 kg m 8 1.601846 x 101 lbm in 8 kg m 3 2.767 990 x 108 Power J
litu h 1 W
2.930 711 x 10-1 Iltu.5-1 W
l.055 056 x 105 ft lbf h-3 W
3.7ff) 161 x 104 ftlbfs-1 W
1.355 818 x 10')
hp (shaft)
W 7,456 999 x 102 Pressure or Stress Ibf in 2 (psi)
Pa 6,894 757 x 105 lbf frr Pa 4.788 026 x 101 kip in 2 Pa 6.894 757 x 106 atmosphere, standard Pa 1.013 250 x 105 inch of mercury (32*F)
Pa 3.38638 x 103 inch of mercury (60"F)
Pa 3.376 85 x 103 inch of water (60*F)
Pa 2.488 4 x 102 11adiation curie liq 3.700 000 x 1010 rad Gy 1.000 000 x 102 rcm Sv 1.000 000 x Ig2 ISJnpIIalutc I
- F
- G tt - (ty - 32)/1.8 "F
K Tg z ' (ty + 459.07)/1.8
- R K
Tx - T.x /1 B I
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N 23A6723 suo. 23 nsv A Tabic 21.1-12 Conversion Factors (Cnntinued)
To convert from seconda ry mtem to primary mtem Multipiv by
.V. Slo tt t v ft min 1 m s*1 5.080 000 x 1&S ft sec-1 mr 3.048 000 x 1&l 1
in sec'l m.s
2.540 000 x 102 rni h-1 m s'l 4.470 400 x 101 Viscosity 2
Ibf sec fr (dynamic)
Pa s 4.788 026 x 101 ft sec 1 (kinematic)
?
m r1 9.290 301 x 1g2 t
volu me ft3 3
2.831685 x 102 m
ins 3
3.638 706 x 103 m
3 8
gal (U.S. liquid) m 3.785 412 x 10 -
Volume per Unit Time ft. min-1 mr1 4.719 474 x IM 3
3 ft sec 1 3
m r1 2.831685 x 192 3
gal (U.S. liquid). min 1 (gprn) m.r 6.309 020 x 105 3 l 2.1.2 Documentation 2.1.2.1 Engineering documentation for the SilWR Standard Plant shall be listed on the Master Parts List (MPL) referenced in paragraph 1.3.1.a. The MPL shall be a controlled list, structured by system, that contains the identification of hardware and software documentation that dellnes the SIlWR Standard Plant. Revision and verification status of the documentation shall be provided.
2.1.2.2 Engineering documentation shall conform to the document standards shown in Table 1.3 2.
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,1 ncv A 2.2 Power Generation Desien 2.2.1 The plant shall be designed to a nominal plant net power of 600 MWe with enndenser circulating water temperature conditions as given in paragraph 2.9 2.2.2 Voltage fluctuation range of the plant generator shall be within i 5%
2.2.3 Normal frequency Ductuations for the plant generator shall be within 58.5 and 60.5 hz with 62 hz maximum allowable for short duration transients. Average frequency over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shall be 60 hz.
2.2A load Following Capability 2.2.4.1 The plant shall be capable of accomodating load demand change: up to i5% at a minimum rate of 5 %/ minute (target rate will be 10 %/ minute).
2.2.4.2 The plant shall be capable of grid frequency regulation within 12 % load change at a -
minimum rate of 0.2 %hecond (target rate will be 1 %/second).
2.2.4.3 The plant shall be capable of daily load following with control rod drive opere. tion between 100% and 50% load on a 14-1-81 hour cycle and with ramp rates up to I %/ minute.
The plant design shall accommodate a minimum of 21900 equivalent daily load following cycles.
2.2.4.4 Power maneuvers within the capabilities above shall not result in isolation or' bypass of condensate /feedwater equipment such as feedwater heaters.
2.2.5 A trip shall not result from a load rejection at 40% or less of rated power.
2.2.6 Plant automation requirements shall be as follows:
Cold startup capability (time interval from initial withdrawal of control rods 'o rated a.
power) shall be less than DATER > hours.
h liot startup capability (time interval from initial load rejection to subsequent t eturn to 90% rated power) shall be less than d ATER> hours.
2.2.7 For plant shutdown purposes, there shall be su!Iicient heat removal capacity a satisfy the i
following reactor coolant temperature reduction schedule:
a.
60.0 *C (140 T) in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
h 54.4 *C (130 *F) in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
]
c.
48.9 *C (120 *F) in 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
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2.2.8 The plant shall accommodate a 24-month refueling interval capability.
2.2.9 'Ihe plant shall be designed to a plant personnel radiation exposure target ofless than 1 man-Sv/ year (100 man-rem / year).
2.3 Safety Dec.icn 2.3.1 U.S. Regulations (para.1.3.4) shall be controlling for design. However, the design shall be adaptable for world-wide application.
2.3.2 The plant design shall comply with the license application information document referenced in paragraph 1.3.1.ac.
2.3.3 The plant shall be designed, constructed, erected and operated in such a way that the release of radioactive material to the emironment does not exceed the limits and guideline values of applicable regulations pertaining to the release of radioactive materials for normal operations and for transients, and accidents (i.e.,10CFR20; 10CRF50, Appendix I; 10CRF100 -
see Table 1.5-1). Transient and accidents are categorized, per NRC Reg. Guide 1.70, as (a)
Incidents of moderate frequency; (b) Infrequent incidents, and (c) Limiting faults.
2.3.4 A pressure suppression type containment that encloses the reactor systems shall be provided. The containment shall function in conjunction with other safety-related features to limit radiological effects of accidents to less than the prescribed acceptable limits (para. 2.3.3).
Piping that penetrates the primary containment and could serve as a path for potential uncontrolled release of radioactive material to the environt shall be automatically isolated when necessary to limit radiological impact to less than acceptable limits.
2.3.5 Industrial sabotage (per 10CFR73) and missiles (per 10CFR50. App. A, GDC 4) originating from outside the plant structuret r. hall be considered, as a design basis. The design basis shallinclude a postulated failure of the turbinesencrator rotor and protection from missiles as required by the document referenced in paragraph 1.3.1.f.
2.3.6 The reactor shall be designed so that nuclear reactivity feedback compensates for rapid power increases for all operating modes (power reactivity coefIicient shall be negative).
2.3.7 The reactor shall be designed so that there is no tendency for divergent oscillation of any operating characteristic considering the interaction of the reactor with other appropriate plant systems.
2.3.8 Safety-related functions shall be provided primarily through passive means. Passive means are natural forces such as gravity and natural circulation, stored energy such as batteries, and compressed fluids, check valves, and non-cycling powered valves.
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2.3.10 Operator actions shall not be required before 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to prevent safety design limits from being exceeded for postulated design buis accident conditions.
2.3.11 The performance of the reactor safety related systems shall satisfy all design criteria when enluated based on up to 105% of rated steam flow.
2.3.12 Safety related actions shall be provided by equipment of sufficient redunda:../ and independence such that no single failure will prevent the required safety-related functions.
2.3.13 Application of the single failure criterion requires that the plant be capable of achiesing (1) emergency reactivity control, (2) emergency core and containment heat removal, and (3) containment isolation and integrity given an inidating event plus an independent single failure in any one of the systems required to support directly or indirectly these safety-related functions.
2.3.14 Multiple failures rest.lting from a single failure are shall be considered to be a single failure.
2.3.15 As a basis for licensing, die core cooling systems shall prmide core cooling in the event of LOCA with pet clad temperature less than 1200*C (2200*F), with a concurent single most limiting failure. As a basis for licensing, the core cooling systems shall prevent core uncovery for any design basis accident with a singic failure.
2.3.16 The design of the plant shall require protection against fires as required by the document referenced in paragraph Ill,h.
2.3.17 The design of the plant shall require protection against flooding as required by the document referenced in paragraph 1.3.1.d.
2.3.18 The design of the plant shtll require protection against external hazard:. of wind and tornadc, and lightning as required by the documents referenced in paragraphs 1.3.Le and 1.3.1.g.
2.5.19 The design shall assure that the safety-related structures. sy:tems, components and equipment shall withstand an operating basis earthquake (Ol3E-refer to paragraph 2.9) in combination with other designated appropriate loads without incurring any damage.
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gr ncy A 2.3.20 The safetyrelated structures, components, equipment and systems for the plant shall be designed to withstand a safe shutdown earthquake (SSE - refer to paragraph 2.9) to assure:
a.
The integrity of the reactor coolant pressure boundary b.
The capability to shutdown the reactor and maintain it in a safe shutdown condition; or c.
The capability to prevent or mitigate the consequences of accidents such that potential release of radioactive material to the emirons does not exceed limits of applicable regulations (10CFR100 - see Table 1.31).
2.3.21 The plant shall have a design basis of withstanding simultaneous and unrelated events as given in Table 2.3-1 while meeting applicable safety criteria, operational requirements and regulations.
2.3.22 The functional classification of sptems as safety-related, special capability or non-safety-related is provided in Table 2.3-2.
2.3.23 Components and equipment of safety-related systems and their supporting systems shall be capable of perfonning their required safety functions under the environmental conditions expected at their locations. Specific requirements are prosided in the specification referenced in paragraph 1.3.1.m.
2.3.24 Credible failures and consequential actions of non-safety.related systems shall not prevent the safety-related systems from accomplishing their safety objectives.
2.3.25 Following a reactor isolation, the plant shall be designed to reach a safe stable condition from reactor critical conditions at rated temperature and pressure within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using safety-related systems and equipment. A safe stable condition shall be attained when the reactor maintains a temperature below 150 C (302*F), the Containment maintains a pressure below design pressure, and combustible gases are controlled below levels that prevents any combustible mixtures in the Centainment spaces.
2.3.26 Equipment in systems designed for emergency core cooling following a LOCA shall be capable of providing continuous system operation with no operator action for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without maintenance, followed by a period of continuous system operation of 100 days with only minor rnf atenance, in addition to in-service periodic tests and applicable transient and shutdown cooling requirements.
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? Notes SSE and OBE are not assumed to be an initiating event of LOCA. However. SSE should be combined with 1
long term LOCA loads, e.g. thermal pressure, etc.
2.
As an initiating event of a transient only.
3.
Mitigation erluipment is seismically qualified, seismic event is considered prior to the fire.
4.
See para. 2.9 for SSE and OBE definition.
D%
5.
Loss of all on-site power, except AC power supplied by uainterruptable power supplies, and ofFsite AC Power is the design basis definition of station blackout.
6.
Transients shall be analped without assuming a coincident SAF, however, transients with a coincident SAF shall be analyzed to determine if the core is covered.
7.
Fire protection system shall be designed to function with LOPA to increase its reliability. Other sptem assumed to have normal power supply available.
8.
Simultaneous occurrence of events shall be defined as those can be considered to occur coincident with the initiating event.
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23A6723 sa No. 30 nrv A I
Table 2.3-2 Functional ClassiGeation of Systems Design Basis j
System Notes Safety-Related Systems El1 Reactor Pressure Vessel System 1,1
]
B21 Nuclear Boiler System 1,2,8,9
-l B32 Isolation Condenser System 1
C12 Control Rod Drive System 11 C21 Leak Detection and Isolation System 2
C51 Neutron Monitoring System C61 Remote Shutdown System C62 Multiplexing System 9
C71 Reactor Protection System C74 Safety System Logic & Control D11 Process Radiation Monitoring System 9
D23 Containment Atmospheric Moniu> ring 5
E50 Gravity-Driven Cooling System i
F15 Refueling Equipment 3
F16 Fuel Storage Facility
]
Jll Nuclear Fuel
]
P52 Instrument Air System 10 R31 Raceway System 6,9 R42 Direct Current Power Supply 6,9 R46 Vital AC Power Supply 6,9 T10 Containment System Til Containment Vessel T12 Containment Internal Structures T15 Passive Containment Cooling System T31 Containment Atmospheric Control System 5
Notp: Designated systems have components which perform the following safety functions:
1.
Pressure Retaining or RCPB function only 2.
Containment isolation function only 3.
Refueling interlocks only 4.
Internal structures with safety-related functions 5.
Ilydrogen monitoring or control function only i
6.
Safety-related equipment only 7.
PRMS used to isolate containment only 8.
Overpressure protection function only 9.
Safety-related instrumentation only
- 10. Supporting safety-relater' crguipment only
- 11. Control Rml Drive System hydraulic scram function only
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Table 2.3-2 Functional Claulfication of Systems Desien Basis (Continued)
System Notes Suecial Capability Systems Cll Rod Control and Information System C41 Standby Liquid Control System C61 Remote Shutdown System P91 Post Accident Sampling System R51 Communication System R52 Lighting and Servicing Power Supply U43 Fire Protection System Non-Saferv-Related Systems CSI Feedwater Control System C82 Automatic Power Regulator System C85 Steam Bypass and Pressure Regulator System C91 Process Computer D21 Area Radiation Monitoring System F42 Fuel Transfer System G21 Fuel & Auxiliary Pools Cooling System G31 Reactor Water Cleanup / Shutdown Cooling System H11 Main Control Room Panels H14 Radwaste Control Room Pancis H21 Local Panels & Racks K10 Liquid Waste Systems K20 Solid Waste Systems K30 OITgas System N11 Turbine Main Steam System N21 Condensate & Feedwater System N22 Heater Drain and Vent System N25 Condensate Purification System N31 Main Turbine N32 Turbine Ccmtrol System N33 Turbine Gland Seal System N34 Turbine Lubricating Oil System N35 Moisture Separator Reheater NSG Extraction System N37 Turbine Bypass System j
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System Notes Non-Saferv-Related Systems (Continued)
N39 Turbine Auxiliary Steam System N41 Generator N42 Hydrogen Gas Cooling System N43 Generator Cooling System N44 Generator Scaling Oil Sprem N51 Exciter N61 Main Condenser and Auxiliaries N71 Circulating Water System P10 Makeup Water System P21 Reactor Building Cooling Water Spiem P22 Turbine Building Cooling Water System P25 Chilled Water System P30 Condensate Storage and Transfer System P32 Oxygen Injection System P33 Process Sampling System P41 Reactor Building Service Water System P42 Turbine Building Service Water Sptem P51 Service Air Sptem P54 High Pressure Nitrogen Gas Supply System P62 Auxiliary Boiler System P73 Hydrogen Water Chemistry System P95 Iron Injection System R10 Electrical Power Distribution System R24 Motor Control Center R31 Raceway System R40 Standby AC Power Supply R42 Direct Current Power Supply R46 Vital AC Power Supply R47 Instrument and Control Power Supply T41 Drywell Cooling System U39 Turbine Building IIVAC U40 Reactor Building HVAC U41 Other Building HVAC U42 Potable and Sanitary Water System U50 Equipment and Moor Drain System j
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33 mv A 2.4 Severe Accident Mitication 2.4.1 Containment pressure during a severe accident shall be limited to the sersice level C capability for the steel pressure vessel and Factored I nad Category for the RCCV based on the -
design pressures listed m the appended data sheet to this specification (para. l.S.S.a).
2.4.2 The containment design leak rate shall be limited to less than 0.5 volume percent per day at rated design pressure, excluding leakage through the MSIVs. MSIV leakage shall be limited to <llster>.
2.4.3 The design shall be capable of accommodating 100% active clad oxidation while maintaining hydrogen concentration to less than 137o or oxygen concentration to less than l
4% in the containment.
2.4.4 Capability to flood the lower drywell cavity shall be provided (para. 3.4.1.3) (to limit production of noncondensibles resulting from interaction of molten core debris with concrete.
Tne cavity floor configuration shall spread the core debris to proside a sumeiently coolable geometry, with an area not less than 0.02 m2 (0.22 ft ) per megawatt rated thermal reactor 2
power. Drywell sumps shall be designed and arranged to prohibit potentiallocalized core debris bed depths in excess of coolable geometry.
2.4.5 The lower drywell floor and walls up to a minimum height shall be lined with, a special material which minimizes gas evolution during inte stion with molten core materials. Height of the wall surfaces with this material shall be sumcient to protect the Containment pressure boundary from failure by contact with the melted core materials (100%
of core) generated during a severe accident. The thickness of this liner material shall be sufTicient to prevent penetration of the molten core material to the pressure boundary during any severe accident events.
5 2.4.6 A core damage Eequency ofless that 1.0 x 10 per reactor year shall be demonstrated by Probabilistic Risk Assessment (PRA).
6 2.4.7 A severe accident frequency ofless than 1.0 x 10 per reactor year for sequences resuldng in greater than 10 mSv (I rem) dose at 805 m (0.5 mile) from the reactor shall be demonstrated by PRA. The source term shall use realistic particulate fraction, credit for scrubbing and delayed fission product release. Off-site dose evaluations shall consider realistic mixing and holdup in the reactor building.
2.5 Station markour Capability 2.5.1 The plant shall be capable of achieving safe reactor shutdown and maintaining shutdown cooling under a postulated loss of all AC power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no operator action.
Concurrently, the DC hatteries shall be designed to be capable of functioning throughout this r,t a smnu na; i
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2.5.2 Startup shall be immediately possibic (no equipment damage or inspection required) when AC power is restored within two hours.
2.6 ATWS Provisions 2.6.1 The following ATWS prevention and mitigation functions shall be provided:
a.
An alternate rod insertion (ARI) consisting of redundant scram air header exhaust valves.
h Fine motion control rod drive (FMCRD) electrical run-in mode for insertion of control blades.
Automatic injection of boron from the Standby Uquid Control Systern (SLCS).
c.
d.
Automatic runback of feedwater flow.
2.6.2 The initiation signals for the ATWS prevention and mitigation functions shall be as defined in the document referenced in paragraph 1.3.1.q.
2.7 Man-Machine Interfaces (MMI) 2.7.1 Building arrangements, equipment locations and instrumentation and controls design shall give primary consideration to the human factors requirements for accessibility, maintenance, operability, contarnination and generation of radwaste, and overall integrated performance.
2.7.2 Control of the reactor shall be provided in a centralized control complex which permits one operator to perform normal operations of the plant. The design shall be premised on a level of operator qualification skill commensurate with personnel possessing an NRC Reactor Operator (RO) license.
2.7.3 Controls, including alarms and displays, shall be arranged to allow the operator to rapidly assess the condition of the plant and locate sptem malfunctions. A program plan shall be developed and implemented to evaluate the effectiveness of the MMI design through ost of mockups, dynam.c simulation and participation of operators as appropriate, 2.7.4 SufEcient indications shall be provided to allow confirmation that the reactor is operating within the envelope of conditions proven safe by plant analysis, NF O 807(Rt V 433)
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2.7.6 Control equipment shall be provided to allow the reactor to resoond automatically to load changes and abnormal operational transients, including initiation of required safety-related
)
actions.
2.7.7 Class 1E qualified display and control devices shall be provided for interfacing with safety-related systems.
1 2.7.8 Interlocks or other automatic equipment shall be provided as backup to procedural j
control to avoid conditions requiring the functioning of plant protection systems or engineered safeguard features.
l 2.7.9 Means to manually control reactor power level shall be provided.
2.7.10 Automation of plant startup, shutdown and post-transient reconfiguration operations shall be provided.
2.7.11 Capability to bring the core to cold shutdown conditions from outside the centralized control complex after reactor scram has been initiated shall be prosided.
2.7.12 Post accident monitoring (PAM) requirements are provided in the appended data sheet (para.1.3.3.a) to this specification.
2.8 Plant Conficuration and Secudtv 2.8.1 The overall plant design shall be for a single unit.
j 2.8.2 The principle structures of the plant shallinclude:
Reactor Building - houses the safety envelope with containment and reactor, all safety-a.
related systems, refueling area with spent fuel storage, the control area, auxiliary area, senice area, senice water area, and liquid waste processing area.
b.
Turbine Building - houses equipment a.w>ciated with the Main Tuthine, Main Condenser and Generator and their auxiliary systems and equipment, including the Condensate Purification System.
2.8.3 All structures, components, equipment and systems providing safety-related functions shall be located within the reactor building.
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2.8.4 Fuel handling and storage facilities shall be designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel as necessary to meet operating and off-site dose constraints. The safety-related heat removal function for spent fuel shall be by providing sufficient water in the spent fuel storage pool so that boil.off of water for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an accident does not exceed the volume that would allow the pool level to come within some minimum depth above the top of the spent fuel. This minimum depth above top of the spent fuel shall be sufficient to provide adequate shielding to allow recovery operations in the spent fuel pool areas.
2.8.5 The spent fuel storage pool shall have a wet storage capacity of 10 years of operation plus one core load.
2.8.6 Non-safety related functions (systems or equipment) which provide backup to safety-related functions shall be physically separated from the safety-related functions. In so far as possible, the arrangement shall minimize the possibility of compromizing the functionality of both the safety related functions and their backup systems and equipment for any credible events which could cause damage to one region of the plant.
2.8.7 The Main Control Room shall be shielded against normal sources of radiadon.
Habitability contingencies shall be prosided so that continued occupancy for a period of time consistent with station blackout requirements (para. 2.5) is possible under accident conditions, including severe accidents.
2.8.8 The overalllayout of the plant shall provide passageways for personnel and equipment in and between the reactor building, turbine building and service building. The access shall be based on separate passageways for clean and controlled areas in all buildings.
2.S.9 Access control measures, building design, equipment layout, equipment design, equipment selection and shielding shall be used to ensure that radiation protection standards can be satisGed in any mode of normal plant operations. The IIVAC systems and equipment shall designed so that airborne radioactive material can be minimized.
2.8.10 To control and minimize personnel exposure, the following design criteria shall apply when applicable and practical to accessible areas:
Nonradioactive equipment requiring access shall not be located in areas with radioactive a.
equipment.
h Radioactive equipment shall be designed tc minimiic crud buildup and to provide for easy decontamination and maintenance.
Radioactive piping should be accessible and shall be arranged to allow inspection and c.
maintenance with minimum exposure but shall not be located so that it causes exposurc to personnel performing normal maintenance.
d.
Personnel access to radiation areas shall be minimized.
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37 REV A 2.8.11 The overall plant arrangement shall be consistent with security requirements for controlling the access of personnel and vehicles. This shall include perimeter fencing, entry guard house and controlled access to the reactor building enclosing safety-related equipment.
Entrance to such buildings for personnel and material s! all be controlled with all external entrances h>cked except during periods of use.
2.8.12 The DC lighting systems shall be designed to provide access to all areas in which manual or planned actions are required during emergency operations and shall remain functional for a period of time consistent with station blackout requirements (para. 2J5). In addition, the emergetty communication system shall also be designed to be functional for this period of time and shall not interfere with instrumentation and control systems.
2.9 Site Environmental Characteristics. See Table 2.9-1.
2.10 Desien. Construction and Transnortability
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2.10.1 Matetinhard Processes. The materials and processes to be used in the design shall be in compliance with the requirements of the materials and processes control document referenced in paragraph 1.3.1.o. Materials and processes shall be specifically selected to reduce potential for intergranular stres, corrosion cracking (IGSCC) and irradiation assisted stress corcosion cracking (IASCC), and to support the personnel radiation exposure target requirement of paragraph 2.2.9.
2.10.2 Mechanical and Structural Loadine 2.10.2.1 Safety-related structures, systems. components and equipment shall be protec,ted against the dynamic loading resulting from an instantaneous severance of those high energy pipes of sizes such that early leakage cannot be detected before breaks. Design requirements are detailed in the document referenced in paragraph 1.3.1.i.
2.10.2.2 Safety-related structures, systems, components ar,d equipment shall be designed to perform their intended safety-related functions following the impingement of a fluid jet resulting from a pipe break. Protective measures to be considered include physical separation and shielding. Design requirements are detailed in the document referenced in paragraph 1.3.1.j.
2.10.2.3 Safety.related structures, systems, components and equipment shall be designed for missiles per design requirements that are specified in the document referenced in paragraph 1.3.1.f.
2.10.2A The dynamic and static loads to be considered in the Reactor Building structural design shall include those load combinations defined in the document referenced paragraph 1.3.1.n.
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Maximum Ground Water Level 0.61 m (2 ft) below grade i
b.
Maximum Mood (or Tsunami) Level 0.30 m (1 ft) below grade or less (note 1) c.
Precinitation (for roofdesien):
- Max. rainfall rate 49.3 cm/h (19.4 in/h) or 15.7 cm(6.2 in) in 5 min 2
- Max snow load 2.394 kPa (50 lbf/ft )
d.
Desien Temtseratures:
- Ambient 1% exceedance values:
Maximum 37.8'C (100 F) dry bulb /25.0 C (77'F) coincident wet bulb Minimum
- 23.3*C (- 10 F) 0% exceedance values (historical limit):
Maximum 46.1*C (115'F) dry bulb /27.8'C (82 F) coincident wet bulb Minimutn
- 40.0 C (- 40*F)
- Auxiliary system and IC cooling water inlet 35.0*C (95 F)
- Main Condenser cooling water inlet s 37.8*C (100*F) c.
Extreme Wind
- Basic wind speed 44.7 m/s (100 mi/h) (note 2)
- Importance factors 1.0 (note 3)/1.11 (note 4) f.
Tornado (note 5)
- Maximum tornado wind speed 116.2 m/s (260 mi/h)
- Translational velocity 25.5 m/s (57 mi/h)
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- Radius 138 m (453 ft)
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- Maximum atmospheric AP 10.1 kPa (1.46 psi) j
- Rate of pressure change 1.86 kPa/s (0.27 psi /s)
- Missile spectra Spectra I of SRP 3.5.1.4
- Missile velocity 35% of Max. Horizontal Windspeed
- Missile altitude 9.1 m (30 ft) above grade for large soft and large rigid missles; all elevations for the small rigid missles NE O fc' $E v dise)
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23A6723 sn o. '39 REV A Table 2.9-1 Envelone of Plant Sit: Desien Parameters (Continuedl Param eter g.
Seismology
- OBE peak ground acceleration 0.10 g (note 6)
- SSE peak ground acceleration 0.30 g (notes 6 & 7)
- SSE response spectra per US NRC Reg. Guide 1.60
- SSE time history Envelope SSE response spectra h.
Soil Properties (note 8)
- Minimum bearing capacity (demand) 718 kPa (15 kip /ft )
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- Minimum shear wave velocity 3M.8 m/s (1000 ft/s)
- lj uefaction potential None (at plant site resulting from q
Noter (1) Probable maximum Good level (PMF), as defined in ANSI /ANS 2.8 " Determining Design Basis Flooding at Power Reactor Sites." Minimum value to be basis of standard plant design with provisions for accommodation of Good levels up to maximum value.
(2) 50 year recurrence interval i
(3) Importance factor to be used for design of non4afety-related structures only as defined in ANSI A58.1.
(4) Importance factor to be used for design of safety-rclated structurcs only as defined in ANSI A58.1.
(5) 1,000,000-year tornado recurrence-interval, with associated parameters based on ANSI /ANS 2.3. Pressure effects associated with potential off-site explosions are assumed to be non<ontrolling for the design.
(6) Free-fic1d, at plant grade evaluation.
(7) Envelopes all present U.S. nuclear sites, except those on California coastline.
i (8) Values ofIraring capacity and shear wave velocity are included in this table to assure wide application of a standard mat-type foundation design. Design must be evaluated parametrically against ranges of possible soil properties to verify wide application.
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2.10.4 Electrical Design Requirements. Electrical design for safety-related systems shall comply with the separation requirements referenced in paragraph 1.3.1.1.
2.10.5 Instrumentation and Control Desien Reanirements The instrumentation and control design process shall follow the requirements of the document referenced in paragraph 1.3.1 aa.
2.10.6 Transoortability Requirements 2.10.6.1 The plant equipment and structural modules shall be designed for maximum use of oft-site shop fabrication. Equipment and structural modules shall be of sizes capable of rail transport to the site to the maximum extent possible. Rail shippable modules must fit in a size envelope of 3.81 x 3.81 x 18.3 m (12.5 x 12.5 x 60 ft). These rail shippable modules can be assembled and integrated into larger modules in the pre-assembly areas on the plant site before placement into the final plant position. Special considerations will be given to extremely large components such as the Reactor Pressure Vessel with evaluations of overland transport versus final site assembly of vessel sub-parts.
2.10.6.2 Dynamic acceleration of plant equipment and structural modules during shipping shall be limited to 3 g in the direction of travel and 1 g in other directions.
2.11 Plant Duty Cycle and Iife 2.11.1 Plant structures, systems, components and equipment shall be designed for the number of thermal and pressure duty cycles expected during the plant lifetime (para. 2.11.3),
including those from normal operation, abnormal transients and design basis accidents.
Design duty cycles are given in the specification referenced in paragraph 1.3.1.t.
2.11.2 Equipment in sptems required for continuous power generation and equipment required for shutdown cooling, refueling, inspection and maintenance, or hot standby conditions shall be designed for an operating life based on continuous plant operation at full power for a 24-month cycle, including a refueling outage of 50 days, and then restart.
2.11.3 The plant design life shall be f.0 years from the full power operating license date with consideration for planned comprment maintenance and replacement. The design availability requirement of paragraph 2.15.2 shall be maintained with consideration of component replacement.
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2.12.2 Plant general requirements (A-series) shall be applied to the various plant system designs as indicated in the document referenced in paragraph 1.3.1.c.
2.12.3 Inidating parameters for protective and engineered safeguard functions are given in Table 2.12-1.
2.13 Testabilitv of Systems and Components 2.13.1 Systems, components and equipment which perform safetyrelated functions shall be designed to permit testing and inspection under conditions of normal operation, or during planned reactor outages for refueling and inspection, to confirm their integrity and capability and to assure that plant design availability requirement defined in paragraph 2.15.2 is met.
2.13.2 Control logic and data acquisition equipment for safety-related functions shall include diagnostics and on-line self-test to detect defects and facilitate repair, as is practical, without inDuencing the correctness of the output functions.
2.13.3 The protection and engineered safeguard systems shall be designed so that:
The plant can be operated indefinitely at full power with one channel in test or bypassed, a.
b.
One subsequent single failure will not cause a plant trip.
2.14 Ouality Assurance. The quality assurance plan described in the document referenced in paragraph 1.3.1.ad shall be implemented to ensure an acceptable level of confidence that the structures, systems, components and equipment will perform required safety-related functions.
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High Neutron Mux High Thermal Power 1(11)
Main Steamline High 1(1)
Ambient Temperature Main Steamline High Flow I(l) i Ifigh Reactor Pressure I
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High Drywell Pressure I
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- Level 2 (B)
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- Level 1 I
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Water Line Pressure Low Manual Scram I
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Pressure Isolation Relief Signal Scram (2) Containmen t System SRV ADS GDCS IC Leak Detection I
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(Individual System)
Turbine Inlet Iow Pressure 1(1)
Main Condenser Low I(1)(9)
Vacuum Turbine Area Steamline 1(1)
High Ambient Temperature Iligh Suppr. Pool Temp.
I(12)
Notes:
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I Initiate function T Terminate function (1) Close MSIV (2) Ilydraulic scram is followed by rod insertion via the electrical drive mechanism (3) Close RWCU/SDC shutdown cooling isolation valves, etc, except MSI\\"s (4) Mode switch in Startup or Run, or. mode switch not in Startup or Run and Trip not bypassed (5) Mode switch other than Run mode (6) Mode switch in Run (7) Trip Main Turbine, runback feedwater flow, and trip CRD flow (8) Also initiate CRD flow injection (9) Trip Main Turbine and Bypass valve closure at separate setpoints (10) Trip feedwater pumps (11) Calculated as part of the NMS functions (12) Initiation of suppression pool cooling i
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arv A 2.15 Reliability. Availability and Maintainability
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2.15.1 The plant systems and equipment shall be designed to rnect the Reliability, Availability and Maintainability (RM,1) requirements of the document referenced in paragraph 1.3.1.z. The plant systems and equipment shall also be designed to meet the rcquirements to facilitate operation and maintenance of the document referenced in paragraph 1.3.1.ab.
t 2.15.2 The plant shall have a design availability of 87% or greater.
2.15.3 The plant shall be designed to a forced outage duration tan;et of 5 days or less per year.
and a major outage target of 180 days or less every ten years.
2.15.4 The plant shall be designed to a target of one or fewer unplanned automatic scrams per year.
2.15.5 The periodic refueling and inspection outage shall have a standard duration of 50 days or less. Critical path activities for refueling shall be 17 days or less.
2.16 Control Room Habitability 2.16.1 Means shall be provided to passively maintain habitable conditions in the emergency operating areas in the event of:
Licensing design basis event plus a.
b.
A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> loss of all AC power sources (i.e., loss of off site power plus loss of the non-safety-related diesel generators). Power from the Vital (Uninterruptibic) AC Power Supply (R46), which has its source of power in batteries,is not included in loss of AC power.
The emergency operating areas are defined as the Main Control Room and other areas where access by the operators may be required during the postulated emergency event.
2.16.2 The Main Control Room complex together with the passive habitability design features shall be designed to ensure that the temperature rise does not exceed - 9.4 C (15'F) during the postulated emergency event described in paragraph 2.16.1.
2.16.3 The Main Control Room complex together with the passive habitability design features shall be designed to ensure that the Main Control ' Room operators will not receive radiation exposures in excess of 50 mSv (5 rem) whole body during the postulated emergency event described in paragraph 2.16.1.
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2.16.3.1 This requirement shall be satisfied without the need for individual breathing apparatus and special protective clothing.
2.16.3.2 Breathing apparatus and protective clothing may be used,if necessary, to protect the occupants of other facilities, such as the Technical Support Center (TSC) and Operational Support Center (OSC).
2.17 Classification of Structures Comoonents. Parts. and Svstems. Structures, components, parts, and systems shall be classified with respect to functional requirements and structural integrity to ensure that the significance to plant safety of a given component can be identified during design, construction, and operation. The relationship between the various classification systems is shown on Table 2.17-1 2.17.1 Safety Clanification. Structures, components, parts, and sptems shall be classified with respect to its nuclear safety function as defined in ANSI /ANS-52.1. Structures, components, parts, and sptems shall be designated Safety Class 1 (SGI), Safety Class 2 (SC-2),
Safety Class 3 (SGS), or Non-Nuclear Safety (NNS) based on its intended function. SG1, SC2, and SG3 are classifications of safety-related structures, components, parts, and systems. The term Non-Nuclear Safety (NNS) shall be synonomous with the term non-safety-related, though in some cases NNS structures, components, parts, and systems may be classified as special capability. In any mode of failure structures, components, parts, and systems designated NNS shall not cause the loss of safety-related function of SG 1, SC-2, or SG3 structurcs, components, parts, and systems.
2.17.2 Ouality Group Classification. Structures, components, parts, and systems containing water, steam, or radioactive material shall classified by Quality Group A, B, C, or D as defined in RC 1.26.
2.17.3 Seismic Clauification. Structures, components, parts, and systems shall be classified with respect to the requirements to withstand the efTects of a Safe Shutdown Earthquake.
Structures, components, parts, and sptems shall be designated as Seismic Category 1,II, or III.
2.17.3.1 Seismic CategoryI shall include all structures, components, parts, and systems which are classified as Safety Class 1,2, or 3 and also includes spent fuel storage structures and all fuel racks. Seismic CategoryI structures, components, parts, and systems shall withstand the effects of an SSE.
2.17.3.2 Seismic Category 11 shall include all structures, components, parts, and systems which perform no safetyrelated functions but whose structural failure or interaction could degrade the functioning of a Seismic Category I structure, equipment, or system to an unacceptabic safety level, or could result in incapacitating injury to occupants of the Main Control Room. Seismic CategoryII structures, components, parts, and systems shall structurally withstand the effects of an SSE.
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2.17.4 Electrical ClassiReation. Electrical structures, components, parts, and systems shall be classified according to IEEE requirements as either Class 1E or Non-Class 1E as defined in IEEE 308. Class 1E equipment are classified as Safety Class 3 (SC-3). All safety-related electrical structures, components, parts, and systems shall be Class 1E. Non-Class IE related to all electrical structures, components, parts, and systems not classified as Class 1E.
2.17.5 Classification Related to Ouality Assurance. Structures, components, parts, and systems shall be classified with respect to the application of quality assurance requirements as defined in 10CFR50 Appendix II. These requirements shall be applied to safety-related structures, components, parts, and systems and are designated as "Q". Other structures, components, parts, and systems may have some 10CFR50 Appendix B requirements applied to them as necessary and are designated as "S".
Structures, components, parts, and systems which have no 10CFR50 Appendix Il quality assurance requirements shall be designated as "N".
Tabic 2.17-1 Classification of Structures. Components. Parts, and Systems Safety-Related, Non-Safety-Related, Non-Safety-Related
& Special Capability Safety-Related Special Capability Safety Class SC1 SC 2 SC3 NNS NNS Quality Group A
11 C
C or D C or D Electrical Classific'ation Cl. IE Non-Cl. IE Non-Cl.1 E Seismic Classification I
11 or Ill III Quality Assurance (10CFR50 App. II)
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CENuc/esEnerg 23A6723 su No. 47 REV A
- 3. REQUIREMENTS FOR SiSTEMS 3.1 Nuclear Steam Suoply Systems 3.1.1 Reactor Pressure Vessel (B1li 3.1.1.1 The Reactor Pressure Vessel (RPV) and internals shall generate, separate and dry steam for direct use in the turbine-generator unit.
3.1.1.2 Steam shall contain no more than 0.1% by weight carryover of moisture at the RPV steam nozzles during normal plant operations.
3.1.1.3 The steam separator and dryer assembly shall be removed from the top of the chimney during refueling.
3.1.1.4 Provisions shall be provided to plug the steamlines and stub lines prior to flooding the RPV and reactor cavity for refueling activities.
3.1.1.5 The RPV shall incorporate a chimney to achieve natural recirculation of reactor coolant with sufIicient flow rate to support continuous operadon at rated power.
3.1.1.6 The RPV shall be designed and fabricated in accordance with applicable codes (para.1.3.5.a) for the design pressure listed in the appended data sheet (para.1.3.3.a) to this specification.
3.1.1.7 The RPV shall be fabricated oflow alloy stcel and clad internally with stainless steel or NiCr-Fe alloy (except for the top head, nozzles and nozzle weld zones which are unclad).
3.1.1.8 RPV Temocrature Restrictions for Occration and Hydrotesting (Later>
3.1.2 Nuclear Boiler m21) 3.1.2.1 The Nuclear Boiler System (N15S) shall provide at the rated power condidon the steam flow capacity and conditions as defined in the reactor heat balance referenced in paragraph
- 1. 3.1.w.
3.1.2.2 Main steam line sizing, flow limiter, configura6an and component selection shall be compatible with a rated dome pressure of 7.171 MPa absolute (1040 psia) and a turbine admission pressure of 6.791 MPa absolute (985 psia). Addidonal requirements are specified in the document referenced in paragraph 1.3.1.p.
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48 nsv A 3.1.2.3 Main' steam line piping volume, capacity and performance of the Safety Relief Valves (SRVs) and depressurization vahes (DPVs), and closure time of the Main Steam Isolation Valves (MSIVs) shall he as specified in the document referenced in paragraph 1.3.1.q.
3.1.2.4 The NBS, in combination with the isolation condenser (para. 3.1.3.4), shall limit reactor pressure to less than the lowest setpoint of the SRVs upon reactor isolation with scram initiated on MSIV closure when in the reactor Run mode.
3.1.2.5 The NBS shall automatically depressurire the reactor using SRVs in combination with DPVs if low RPV water level is signaled (see Table 2.12-1). Depressurization must be comp!cted in time to allow gravity 4 riven cooling system (GDCS) injection flow to replenish core coolant sufIiciently to satisfy the requirements of paragraph 2.3.15 when considering failure of any single active component. Depressurization to a pressure of <lATER> shall be accomplished with only SRVs and DPVs operating.
3.1.31colation Condenwr m321 3.1.3.1 The Isolation Condenser System (ICS) shall consist of three high pressure reactor isolation condenser loops. Each IC loop consists of a steam condenser with a steam supply line, condensate return line, vent lines, control and isolation valves. The isolation condenser system parts which are located downstream of the isolation utives and outside the containment are not safety-related. The ICs share a common cooling pool (IC/PCCS pool) with the PCCs of the Passive Containment Cooling System (T15). The IC/PCCS poolis safety-related.
3.1.3.2 The ICs receive their steam supply from lines connected directly to the reactor pressure vesscl (PJV). The ICs operate by natural circulation of steam with grasily return of condensate to the reactor vessel, so they require the opening of a condensate drain valves to initiate their operation on high reactor pressure. Since the ICs are connected to the RPV and are located outside the containment, the piping includes containment isolation valves.
3.1.3.3 The ICs shall be sized to remove post-reactor isolation decay heat with 2 out of 3 ICs operating without reactor overpressure (the ICs are not sized to reduce reactor pressure and temperature to cold shutdown conditions).
3.1.3.4 The ICS shalllimit reactor pressure to less than the lowest setpoint of the SRVs for moderately frequent events resultingin reactor isolation as described in the requirement of paragraph 3.1.2.4. Furthermore, the ICs, together with the water stored in the RPV, shall conserve suflicient reactor water to avoid automatic depressuriration from low reactor water level (Table 2.12-1).
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3.2.1 Rod Controf andInformation (Cl1) 5.2.1.1 The Rod Control and information S stem (RC&lS) shall be a fault tolerant system to 3
control the Fine Motion Control Rod Drive (FMCRD) motors of the Control Rod Drive System, and shall acquire, supply, and display information about positions of the control rods in the core and status of FMCRDs to the operator and other plant systems.
3.2.1.2 nrough the capabilities of gang rod selection and verification logic, the RC&lS shall enforce adherence to a predetermined rod pull / insert sequence as required by the Rod Worth Minimizer subsystem and automation interface with the Power Generation Control System (PGCS), part of Process Computer (C91), (para. 3.2.13).
3.2.1.3 RC&lS shall support the load following requirements of paragraph 2.2.4, and the automation requirement of paragraph 2.2.6 through its automation interface with the Power Generation Control System (PGCS) (para. 3.2.13) and the Automatic Power Regulator (para.
3.2.11).
t 3.2.1.4 The RC&IS shall direct rod insertion to full-in position by electrical power upon receipt of a signal from the Reactor Protection System indicative of reactor trip (scram) initiation, or receipt of an ATWS related signal as required in paragraph 2.6.
3.2.1.5 The RC&lS shall prevent potentially unsafe rod movements by automatically enforcing rod movement blocks and, when in automatic power control mode, by automatically monitoring fuel operating thennal limits. Rese blocks shall not impact the hydraulic scram function, the scram following function-or ATWS mitigation functions.
3.2.2 Control Rod Drive (C12) 3.2.2.1 A Control Rod Drive (CRD) System with a Fine Motion Crmtrol Rod Drive (FMCRD) shall be provided to position neutron alwirbing control rods by electrical means within the core in response to comrnands from the RC&lS for power manipulations and to rapidly insert the control rods (scram) by hydraulic means for the safe shutdown of the reactor during anticipated or abnormal operating transients.
I 3.2.2.2 ne FMCRD shall provide electrical control rod insertion for any anticipated event when either normal and/or standby plant power supplies are available, even if normal hydraulic scram is assumed to fail.
3.2.2.3 Control rod motion shall be limited to values in accordance with the document referenced in paragraph 1.3.1.s in the event of Control Rod Drive housing failure. This shall be accomplished without external supports.
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3.2.2A The BfCRD shall have features which limit control rod motion,in the event of a i
scram line break, to values in accordance with the document referenced in paragraph 1.3.1.s.
l 3.2.2.5 Dual Class IE separation switches shall be provided on the BtCRD to block rod j
withdrawal when control rod separation is detected.
l 3.2.2.6 The RfCRD shall function with either a Il C control blade, a hafnium control blade, 4
or an equivalent blade of approwd material.
3.2.2.7 The Control Rod Drive (CRD) sptem shall use ganged scram accumulators with two drives per hydraulic control unit. During a scram, water from the accumulators shall be discharged direcdy into the RfCRDs and, subsequently, into the RPV. There shall be no external scram discharge volume.
3.2.2.8 The CRD system shall provide high pressure makeup capability as defined in the document referenced in para.1.3.3.a, to the RI%ia a bypass line connecting the CRD pump discharge header to the RWCU/SDC piping and hence to feedwater inlet piping and through the CRD purge.
3.2.2.9 The control rod scram speed is speciRed in the document referenced in paragraph 1.3.1.s.
3.2.2.10 The BfCRD stepwidth of travel is specified in the appended data sheet (para.1.3.3.a) to this specification.
3.2.2.11 The B1CRD will be supplied with AC porter with backup by on-site standby AC power.
3.2.3 Iral Detection and isniation (C21) 3.2.3.1 The Leak Detection and Isolation System (LD&lS) shall detect, indicate and alarm leakage from the reactor primary pressure boundary and, when required, initiate closure of primary containment isolation valves. The isolation condenser tubing, ofisolation condensers attached to the RPV,is included as part of the primary pressure boundary. The LD&lS shall also isolate individual systems as required.
3.2.3.2 Refer to the appended data sheet (para.13.3.a) to this specification for isolation group classiGcation and containment isoladon drawing.
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3.2.4.2 The feedwater controller and data acquisition system shall employ digital controls which incorporate triplicated-channel fault-tolerant design for active components and provide for self-contained on-line testability and repair.
3.2.4.3 The logic for tripping feedwater system components shall not permit inadvertent tripping, nor prohibit tripping due to a single failure, including loss of power source.
3.2.4.4 A reactor scram on low water level shall be precluded following the loss of one operating feedwater pump or one operating condensate pump from full power operation.
3.2.4.5 During reactor startup automatic level control shall be facilitated by control of the RWCU dump vahc.
3.2.4.6 FWCS shall include provisions for runback of feedwater runback on receipt of an ATWS signal to flow values stated in the document referenced in paragraph 1.3.1.q.
1 3.2.5 Standbv Liould Control (C4J1 3.2.5.1 The etandby liquid control sptem (SLCS) shall provide a separate and diverse means for inserting negative reactivity into the reactor core.
3.2.5.2 SLCS shall be initiated for ATWS mitigation as required in paragraph 2.6.1.c. Injection for SLCS shall be accomplished using an accumulator tank that is maintained at a minimum pressure of 10.34 MPa gauge (1500 psig). Redundant squib actuated injection vahes shall be employed for system actuation.
3.2.6 Neutron Monitorine (C51) 3.2.6.1 A digitalin < ore Neutron Monitoring Sptem (NMS) shall be provided to detect normal and excessive power generation conditions, and by appropriate signals to the Reactor Protection System (para. 3.2.9) shall prmide protection against fuel damage during planned operations, anticipated or abnormal operational transients and accidents.
3.2.6.2 A Startup Range Neutron Monitoring (SRNM) subsystem shall provide a continuous measure of neutron flux and reactor period from source level to approximately 10% of rated power using fixed in-core startup range neutron monitors.
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3.2.6.4 The design shallincorporate an automated sptem for calibrating the Local Power Range Monitors (LPRMs) of the PRNM using fixed in-core gamma thermometer detectors.
3.2.6.5 Scram trip setpoints shall be as given in the documents referenced in paragraphs 1.3.1.q and 1.3.1.s.
3.2.6.6 Detector and source locations are shown in the appended data sheet (para.1.3.3.a) to this specification.
3.2.7 Remote Shutdown (C61) 3.2.7.1 The Remote Shutdown System shall provide control of the reactor systems necessary for prompt controlled hot shutdown with subsequent capability to attain cold shutdown from outside the Main Control Room, in accordance with the MMI requirements of paragraph 2.7.11.
3.2.7.2 The remote shutdown system shall be single failure proof and scismically qualified.
3.2.8 Multiplexing System (C62). A multiplexing sptem using fiber optic main transmission lines shall provide distributed control and instrumentation data communication networks to
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support the monitoring and control ofinterfacing plant systems. Multiplexing shall be applied for both non. safety.related and (separately) safety-related functions.
3.2.9 Reactor Protution (C71) 3.2.9.1 A Reactor Protection System shall be provided to initiate an automatic shutdown via hydraulic insertion of control rods (reactor scram), and electric run in after scram, if monitored system variables exceed preestablished limits during any conditions of normal operations, anticipated or abnormal operational transients and accidents.
3.2.9.2 The initiating parameters for scram shall be as given in Table 2.12-1. Trip setpoints shall include those in the document referenced in paragraph 1.3.1.q.
3.2.10 Safety Svstem Locic & Control (C74) 3.2.10.1 Safety Sptem logic & Control (SSLC) shall provide a centralized facility for implementing safety-relaten Icgic functions. These functions, which are described on logic diagrams of plant r,afety systems or are safetyrelated functions of other plant sptems, enable the safety sys. cms to perform thei: plant protection tasks.
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3.2.10.3 SSLC shall acquire data from redundant sets of sensors of the interfacing systems and provide control outputs to the final component actuators. Data shall be received from the essential multiplexing sptem (part of C62, para. 3.2.8) or directly hardwired from transmitters or sensors. The supported interfacing sptems shall be as follows:
Complete Ix>gic Diagram:
a.
(1) D21 Nuclear Boiler System (2) B32 Isolation Condenscr (3) C21 Leak Detection & Isolation System (4) C71 Reactor Protection System (5) E50 Cravity-Driven Cooling Sptem h
Selected Isolation or Initiation Functions:
(1) P54 High Pressure Nitrogen Supply System (2) T3I Containment Atmospheric Control System (3) T49 Flammability Control System (4) T53 Suppression PoolTemperature Monitoring (5) U40 Reactor Building HVAC (6) U41 Other Building HVAC 3.2.10.4 SSLC shall perform the following functions:
Sensor channel trip decisions a.
h System coincidence trip decisions (2-out-of-4 logic) c.
Control and interlock logie d.
Manual dhision trip and isolation Division-of-sensors bypass j
e.
f.
Division maintenance bypass (division out-of-senice) g.
Calibration and self-diagnosis 3.2.11 Automatic Power Reculator (C82) 3.2.11.1 The. Automatic Power Regulator shall receive off41te load demand requests, compare the requested rates and magnitudes to preestablished limits and provide Control Rod withdrawal and insertion demand signal to RC&lS (Cll) and load demand signal to the Turbine Control System (N32) to accomplish load following.
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i 3.2.12.2 The controller and data acquisition system shall employ digital controls which incorporate triplicated-channel fault-tolerant design for active components and provide for self-contained on4ine testability and repair.
3.2.12.3 SB&PC shall support tha trip avoidance requirement forload rejection events specified in paragraph 2.2.5.
3.2.12A In conjunction with the Automatic Power Regulator (pa..t. 3.2.11) and the Turbine Control System (N32), SB&PC shall support the load following requirements of paragraph E.2.4.
3.2.12.5 The turbine bypass capacity requirement is specified in paragraph 3.10.4.3 3.2.13 Process Computer (C91) 5.2.13.1 The process computer shall provide a Performance Monitoring and Control Syste:n (PMCS) function and Power Generation Control System (PGCS) function to support etlicient plant operation and automation.
3.2.13.2 The PMCS shallinclude:
Calculation of three-dimensional core power distribution.
a.
h Determination of margins to fuel operating limits, Predic.ive determinations of core performance and margins to fuel operating limits c.
aunciated with planned control rmi movements.
d.
Calculation of balance of plant performance.
Displaying and recording of plant wide transducer data and status during startup, e.
normal oper2 tion and both planned and forced reactor shutdown.
f.
Generation of alarm or change of state outputs as required to plant annunciators and to other operator information systems.
g.
Provide CRT display and touch control capability, drive the large displays in the Main Control Room, SPDS, sequence of events recording, plant automation, and emergency
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procedure tracking function (see document referenced in para.1.3.1.aa).
j h.
Calculation of core flow by the heat balance method.
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3.2.13 A The Process Computer shall provide the necessary interfaces with the Technical Support Center (TSC).
3.2.13 A Failure of the Proces.* Computer shall not result in a loss of plant shutdown and monitoring capabilities from the Main Control Room.
3.3 Radiation Monitoring Swtems 3.3.1 Process Radiation Monitorine (D11). The Process Radiation Monitoring System shall monitor the fission products in process and efIluent gas and liquid lines that might serve as discharge paths for radioactive materials, provide alarm on abnormal indications and, as required, initiate actions to contain fission product release.
3.3.2 Area Radiation Monitoring (D?l). The Area Radiation Monitoring System shall monitor the gamma radiation levels in selected plant locations and in areas where nuclear fuel is stored or handled. The sptem shall provide plant personnel with records and indications of the monitored levels and shall warn personnel of excessive radiation levels by initiating alarms in the control room as ' ell as in local areas as required.
w 3.3.3 Containment Atmospheric Monitorine (D23). The Containment Atmospheric j
Monitoring System (CAMS) shall monitor the gamma radiation dose rate as well as the hydrogen and oxygen concentration levels in the drywell and wetwell areas of the containment during normal plant operation, shutdown and during post-accident following a LOCA event. TF.c sptem shall consist of two sedundant and independent divisions and eacn division shall provide records and indications in the control room of the monitored input variables on a sampling bads. The system shall also initiate alarms in the control room to alert the operator of any abnormallevels.
3 A Core Cooling Systems 3A.1 Gravity-Driven Cooling (E50).
3A.l.1 The Gravity-Driven Cooling Sptem (GDCS) shall provide emergency core cooling in conjunction with the automatic depressurization function of the NBS (para. 3.1.2.5) by replenir.hing coolant in the core via gravity head when the RPV is at low pressure.
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dawell pressure (Table 2.12-1).
t 3.4.1.3 The lines from the GDCS pool shall have pipe extensions with thermal actuated deluge ulves terminating in the lower drywell. The valves will actuate to flood the lower drywell to the height specified in paragraph 3.14.2.5 given thermal condidons indicative of a severe accident with molten core material on the drywell floor.
3.4.1.4 There shall be sufIicient initial water inventory in the GDCS pools-to flood the-Containment Vessel (and including the' inner PSV volume and lower drywell) to at least 1.00 m (3.28 ft) above the top of the active fuel (TAF) assuming a single failure.
3.5 Reactor Servicing Eauipment (F-< cries) 3.5.1 The refueling system shall provide for reactor refueling, servicing operations within the RPV and insenice inspection of the RPV and piping. The normal RPV senice operations to be performed are removal and replacement drywell head, RPV head, separators, neutron sources, installation of control rods and flooding and draining of the equipment storage pool.
3.5.2 The fuel senicing funcdons to be performed are receiving, inspection, storage and rechanneling of new fuel. For irradiated fuel the functions to be performed are fuct sipping senices, relocation to different core locations, and eventual discharge to the fuel storage pool.
Provisions shall be made for irradiated fuel to be handled, shuffled, stored and loaded into shipping casks.
3.5.3 A system ofinterlocks that restrict movement of refueling equipment and control rmis shall be provided to prevent an inadvertent criticality during refueling.
3.5.4 Other capabilities shall include senicing of CRD, in< ore instruments, equipment located in the lower drywell and general equipment located in the reactor building.
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and auxiliary pools cooling.
h Provide low pressure coolant injection of suppression pool water into 'he reactor pressure vessel.
c.
Provide low pressure coolant injection of Condensate Storage Tank water into the Reactor Pressure Vessel.
3.6.1.2 FAPCS mode shall provide capability for a minimum of two changcouts per day of the coolant in the spent fuel storage pools and keep the fuel pool maximum temperature at or below 48.9 C (120"F).
3.6.1.3 FAPCS mode shall provide capability for a minimum of two changeouts per day of the collective pool coolant volumes, i.e., the spent fuel storage pools, the fuel transfer pool, the alove reactor pools including isolation condenser pools, the GDCS pools and the suppression pool, while maintaining water quality within the limits of the specification referenced in paragraph 1.3.1.u.
3.6.2 Reactor Water Cleanup / Shutdown Coolinc System (G311 3.6 2.1 The Reactor Water Cleanup / Shutdown Cooling System shall:
a.
Maintain reactor water quality within the limits specified in the document referenced in paragraph 1.3.1.u.
h Transfer sensible heat and core decay heat load produced when the reactor is being shut down, or is in the shutdown condition, to a secondary coolant system.
c.
Provide the capability to overboard water and provide a line that can accept high pressure makeup water from the CRD System (Cl2) so that reactor water level can be controlled by the Feedwater Control System (CSI) during startup and shutdown.
d.
Provide high pressure cooling of the primary coolant during periods of reactor isolation t
(hot standby).
3.6.2.2 The RWCU/SDC System shall consist of a dual train arrangement which allows maintenance of major system equipment of one of the trains while the the other train operates to satisfy functional requirements. For both the RWCU and the SDC modes 100% system capability shall be achieved with both trains operating.
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3.6.2.4 RWCU mode shall provide an adjustable amount of primary coolant water purification up to a minimum capacity of 1.0% flowrate (feedwater mass flowrate equivalent) given any one system train out of serdce. With both system trains in operation the RWCU mode shall provide an adjustable amount of primary coolant water purification up to a minimum capacity of 2.0% flowrate (feedwater mass flowrate equivalent).
3.6.2.5 During normal service in the RWCU mode, RWCU/SDC shall have one train operational while the other train is in standby. Each of the two trains shall have suction from its own RPV connection and its owm bottom head drain connection, and shall discharge separately to one of the two feedwater lines.
3.6.2.6 During plant startup, RWCU/SDC shall provide overboarding of purified reactor swell water volume, together with control rod drive purge water entering the vessei,in quantities suflicient at all times to cope with reactor temperature rises of 37.8 C/h (100*F/h), and during which for any hourly period, RPV net water level increase is held to one<}uarter the normal control range.
3.7 Control Pancis 3.7.1 Main Control Panels (H10). The main control room panels shallinclude a main console of a U-shaped configuration which shall be human engineered to allow one man operation from a seated position for normal operating conditions. In addition, a set oflarge, integrated plant status displays shall be located in the main control room so as to be clearly visible by the control room operating team.
3.7.2 Main Control Ronm (MCR) Enuipment Room Panels (Hil) The Main Control Room (MCR) Fquipment Room Panels shall contain the system controllers, control room multiplering units, and control logic and displap and alarms, as necessary to satisfy the system configurations described in the individual system requirements documents.
3.7.3 Radioactive Waste (Radwaste) Control Room Panels (H14). Radwaste Control Room Panels shall be provided as the primary operator interface for the Liquid and Solid Waste Systems (para. 3.9.1). These control panels shall be designed in accordance with the MMI requirements of paragraph 2.7.
3.7.4 Iecal Pancis and Racks (1121). Local panels or cabinets shall be provided for the electronics to effect system operation and shall be designed for uniformity using rigid steel structure capable of maintaining electrical integrity as required under seismic and plant dynamic conditions. Local acks shall be provided to form mounts and supports for the N E O $07 (REV 4'65;
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3.8 Nuclear Fuel (J-series) 3.8.1 Rated core thermal power of 2000 MW at a reference core pressure of 7.239 MPa absolute (1050 psia) shall be the core and fuel design basis to meet the generator output (para. 2.2.1),
3.8.2 The reactor core shall contain 732 fuel bundles and 177 control blades. The core arrangement is shown in the appended data sheet (para,1.3.3.a) to this specification.
3.8.3 The reference equilibrium cycle basis shall be a reference fuel cycle of the interval specified in paragraph 2.2.8 with an operating capacity factor of 98.5%.
3.8A The reference initial cycle basis shall be a reference fuel cycle of the interval specified in paragraph 2.2.8 with an operating capacity factor of 93.5%.
3.8.5 The design basis reload fuel bundle average discharge burnup shall be 38,000 MW d/ tonne, with future capability consistent with the fuel product line limits. The fuel mechanical design shall accommodate a peak assembly-average burnup of 50,000 MW d/ tonne.
3.8.6 Fuel for the reactor shall consist of slightly enriched uranium dioxide pellets sealed in Zircaloy-2 cladding tubes. The fuel cladding shall be designed in conjunction with other plant systems to retain integrity so that the consequences of failures are within acceptable limits (para. 2.3.3) for the design life of the fuel, throughout the range of normal operational conditions and anticipated and abnormal operational transients.
t 3.8.7 The core shall be designed with at least 15% margins for minimum critical power ratio 1
(MCPR) and maximum linear heat generation rate (MLHGR) relative to operating licensing limits as demonstrated with simulated control rod pattern sequencing through the initial and i
equilibrium cycles. Target operational licensing limits shall be as follows:
a.
MCPR: 1.40 (Evaluated limit will be reported in the document referenced in paragraph 1.3.1.x h
MLHGR: 47.2 kW/m (14.4 kW/ft) 3.8.8 The response to transient conditions shall meet the following stability criteria (applied in the document referenced in para.1.3.1.q):
Core decay ratio: less than or equal to 0.40.
a.
b.
Hot channel decay ratio: less than or equal to 0.30.
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k-eff of 0.99) with a pair of rods on the same HCU fully withdrawn, and with one of the pair of rods of highest reactivity worth at all cycle exposure conditions.
3.9 Radioactive Waste Manacement Sutems 3.9.1 Licuid and Solid Radwaste (K10. K20) 3.9.1.1 The Liquid Radwaste System shall be based on not regenerating resins. The use of concentrated chemical decontamination solutions shall be minimized, but the Liquid Radwaste System shall be compatible with such use. The Liquid Radwaste System shall have storage and treatment capacity to process the maximum expected volume. Liquid effluent released to the environs shall comply with guidelines of 10CFR20 and 10CFR50 (Table 1.5-1).
Liquid efIluent recycled to the condensate storage tanks shall meet condensate water quality requirements as given in the specification referenced in paragraph 1.3.1.u.
S.9.1.2 The Solid Radwaste System shall be simplified to asoid the complexities of incineration and waste solidification. On site storage of processed solid wastes awaiting shipment shall be provided for six months (which may include a maintenance outage). Solid wastes for off-site shipment shall comply with the guidelines of 10CFR71 (Table 1.3-1).
Facilities shall be provided to support packaging of solid wastes by a vendor.
3.9.1.3 The Liquid Radwaste system shall automatically stop liquid discharge upon receiving a trip signal from the Process Radiation Monitoring System (para. 3.3.1).
3.9.1.4 Failures of radwaste equipment which may release the radioactive liquids shall not cause an off-site dose to exceed the guidelines of10CFR100 (Table 1.31). Uncontrolled and unmonitored off-site discharges shall be precluded.
3.9.1.5 Both systems shall be designed to prevent inadvertent release of radioactivity during normal operation and maintenance, shall provide remote operadon of all subsystems, and shall minimize the lengths of slurry bearing lines from the reactor water cleanup, fuel pool cooling and cleanup, and condensate treatment systems.
3.9.2 Offgas (K30) 3.9.2.1 Noncondensible gases from the main condenser shall be collected by the offgas ejector and diluted with steam. The oxygen and hydrogen shall be catalytically recombined.
i Bulk water shall be condensed and treated in the Liquid Radwaste System and returned to the Main Condenser. Dried gas shall be passed through charcoal at ambient temperature and discharged through the Stack (U73).
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3.10 Power Cvele Systems l
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3.10.1 Condensate and Feedwater (N21) 3.10.1.1 The Condensate and Feedwater System shall prodde a minimum transient feedwater flow capacity of 150% NBR. This transient capacity shall be maintained for at least (Later> seconds after a turbine trip / generator load rejection. After <Later> seconds, at least
<Later> NBR steady state capacity is required.
3.10.1.2 The drainage configuration and number of heater stages :n the feedwater system shall conform to the turbine cycle heat balance referenced in paragraph 1.3.1.y.
3.10.1.3 All feedwater heater drains shall be cascaded to the main condenser.
3.10.1.4 The reference 9 tem shall have three 33% to 60% capacity (of total system rated flow) adjustable speed motordriven feedwater pumps with nriable frequency power supplies.
Reactor feedwater pumps shall be provided with separate pump shaft seal and bearing cavities that climinate any possibility of oil contaminating the water and sice versa.
3.10.1.5 The reference system shall have three motor-driven, constant-speed, condensate pumps with pump rated flow of 33% of total system rated flow and 60% runout flow capability.
3.10.1.6 Following a trip of any pump in the Condensate and Fcedwater Eystem, the minimum feedwater anilable to the reactor vessel shall be greater than, or equal to, f00% of rated to prevent reactor scram due to low water level.
3.10.1.7 Means shall be provided to recirculate feedwater from after the last feedwater heater (and after the final feedwater sample point) to the condenser hotwell. This recirculation line shall be operative during shutdown from startup conditions to rated pressure conditions and approximately 5% power, and during hot standby. The recirculation line shall be siicd to carry a flow equal to one-third rated feedwater flow.
3.10.1.8 In order to minimize the thermal cycles on the reactor feedwater sparger and nozzle, the feedwater system shall be designed to allow feedwater heating to commence at turbine -
power of 10%
3.10.1.9 A low flow (5:15% NBR) control valve shall be provided for plant startup from low to full pressure and for adequate control overlap at the starting conditions for the feedwater pu m ps.
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sg ncy A 3.10.1.10 Feedwater flow runout capacity limits are given in the document referenced in paragraph 1.3.1.u.
3.10.1.11 The Condensate and Feedwater System shall be designed to ensure that no single failure will result in a plant trip.
3.10.1.12 No single operator error or active equipment failure shall cause a reduction of more ihr
'A C (100'F) in feedwater temperature.
% 10.1.13 The Condensate and Feedwater System shall be designed to include feedwater temperature reduction capability to facilitate end of fuel cycle coastdown. This feedwater tempcrature control shall be generally smooth without large discrete changes in temperature.
3.10.1.14 Control functions shall be provided to allow automatic reactor power reductions upon isolation of feedwater heaters.
j 3.10.2 Heater Drain and Vent (N22)
<LATER>
3.10.3 Condensate Purification (N25) 3.10.3.1 The condensate treatment system shall be of the hollow fiber filter plus mixed bed type.
3.10.3.2 Condensate water quality shall meet the requirements of the water quality specification referenced in paragraph 1.3.1 u.
3.10.3.3 For plants using cooling water of high chloride concentrations (e.g., seawater), a return flow line to the condenser hotwell shall be provided on the combined effluent from the condensate demineralir.ers. This line shall provide closed cycle recirculation capability of the hotwell condensate via the demineralliers in case of a large condenser tube leak. The flow capacity shall be sufIicient to allow operation of one condensate pump.
3.10.3.4 The basic minimum requirement of the condensate treatment system regarding chloride is that sufficient ion exchange capacity shall be available at all times, to maintain the reactor water chloride concentration below 60 ppb for one hour, with a 50 gallon per minute condenser tube Icak rate of seawater of approximately 20,000 ppm chloride,in the case of seawater cooled plants. Equiv; dent system capability shall be available in all other plants with cooling water other than seaw.ter. A minirnum average ion exchange capacity of 50% of theoretical shall be maintainc at all times during normal operation to maintain ample margin for a major condenser tube leak.
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63 atv A 3.10.4 Main Turbine and Auxiliaries fN31 to N39. Nil) 3.10.4.1 The main turbine-generator shall be a tandem compound,1800 rpm, non-reheat, two flow turbine with 52-inch last-stage buckets capable of supplying the nominal electrical output given in paragraph 2.2.1.
3.10.4.2 Sptem piping shall be sized to be consistent with the pressure drops defined in the turbine cycle heat balance referenced in paragraph 1.3.1.x.
3.10A.3 Turbine bypass capacity shall be <LATER>% of rated steam flow.
3.10A A The Main Turbine and Generator control and protection systems shall be designed such that a single failure will not result in a Main Turbine trip.
3.10A.5 The main turbine control valves and turbine bypass valves shall be designed to be compatible with control functions of the SB&PC (para. 3.2.12).
3.10.4.6 The turbine load control function of the Turbine Control System (N32) shall be a triplicated-channel fault-tolerant digital control system that interfaces with the SB&PC (para. 3.2.12) and the APR (para. 3.2.11). Refer to the SB&PC (para. 3.2.12) for additional requirements on the Turbine Control System.
3.10A.7 The turbine load control function of the Turbine Control Sptem (N32) shall be designed to permit uninterrupted plant operation with one channel out of service for repair or maintenance.
3.10A.8 A Main Turbine trip is initiated upon failure of two of the three channels of the fault-tolerant digital preaure control function of the Turbine Control System (N32). Likewise, the failure of two of the three channels of the turbine load control function of the Turbine Control Sptem (N32) shall initiate a Main Turbine trip.
3.10A.9 The Main Turbine overspeed protection function of the Turbine Control System (N32) shall be designed to accommodate a grid frequency increase profile defined as follows, without actuation of the emergency overspeed protection spiem: <Later>
3.10A.10 The Main Turbine shall be designed to accommodate the reactor startup and shutdown profiles defined as follows: <Later>
3.10.4.11 'Ihe Turbine Control System (N32) shall be designed to allow automation of the normal warmup, acceleration, synchronization, and loading sequences, as well as the normal shutdown sequences. The automation control signals will originate from PGCS function of the Process Computer (C91).
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64 nw A 3.10.4.12 The Main Turbine / Generator control and display functions shall be compatible with CRT display and CRT touch-control, as well as other electronically alterable control and display desices in the Main Control Room.
3.10.4.13 In the course of the design of the Main Turbine and Generator those control and protection signals that are not allowed to be multiplexed and those monitoring signals that may be multipicxed shall be specified.
3.10.4.14 Main Turbine interface with the Reactor Protection Sptem (pani. 3.2.9) will not be required.
3.10.5 Generator and Auxiliaries (N41 to N51) 3.10.5.1 The main generator shall be designed so as to satisfy the rated electrical power given in paragraph 2.2.1 and operate in conjunction with the Main Turbine as specified in paragraph 3.10.4.1. The design shah provide three phases of current at 60 hz with direct water cooling of the stator windings and direct hydrogen cooling of the rotor windings.
3.10.5.2 The auxilliary systems for the Generator shall consist of:
Hydrogen Gas Cooling Sptem (N42) for direct cooling of the rotor and indirect cooling a.
of the stator.
b.
Generator Cooling System (N43) for water cooling of the stator.
c.
Generator Scaling Oil System (N44) for providing shaft seal oil to prevent hydrogen leakage.
d.
Carbon dioxide supply from Hydrogen and Carbon Dioxide Bulk Storage (N45) to purge the hydrogen for maintenance.
Exciter (N51) provides DC power for field excitation of the rotor.
c.
3.10.6 Main Condenser and Auxiliaries (N61) 3.10.6.1 The Main Condenser tube material and tube sheets shall be constructed from materials that are not copper-based and are compatible with the document referenced in paragraph 1.3.1.u.
3.10.6.2 The Main Condenser shall be designed to accommodate turbine bypass steam (para. 3.10.4.3), with provisions to prevent mechanical damage to the Main Condenser.
3.10.6.3 The Main Condenser shall have provisions to allow periodic cleaning of the hotwell during plant shutdown.
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,_ ss REV A 3.10.6A The Main Condenser and circulating water system, and tarbine building arrangement shall be designed to permit maintenance (e.g., plugging of tubes and cleaning of water boxes) while the plant is at power. The design shall include the following prodsions to minimize dose rates:
Routing of extraction steam piping to feedwater heaters shall be on the side opposite from a.
the condenser inlet water box where most maintenance would be performed. Extraction steam piping to feedwater heaters shall not be h>cated in the Main Condenser neck.
h.
Providing a feedwater bypass line around heaters located in the Main Condenser neck in order to stop flow of extraction steam and thus reduce dose rates.
The extraction steam piping should be a senice proven material suitable for the c.
conditions as defined in the document referenced in paragraph 1.3.1.u and resistant to i
corrosion <rosion.
3.10.6.5 Any cold, aerated water stream (such as condenser makeup) shall enter high in the condenser through atomizing devices to assure that the water is heated to saturation and deacrated prior to entering the hotwell.
3.10.6.6 The Main Condenser design shall include provisions to detect tube leakage including:
Divided hotwells. The conductivity of the condensate from each of the hotwell sections a.
shall be monitored in order to quickly localize the source of any leak, b.
Sampling trough t.nder each tulesheet.
3.10.6.7 The design of the condenser and circulating water system shall include provisions to effectively control dissolved oxygen in the condensate at r. art and low-load operating conditions in accordance with the document referenced in paragraph 1.3.1.u.
3.10.6.8 'Ihe Main Condenser hotwell capacity, as a minimum, shall be adequate to provide
<LATER> minutes of rated condensate flow.
3.10.6.9 System interfaces shall le provided for draining the Main Condenser, if neces-ary, to i
the Condensate Storage Tank or Suppression Pool when water quality allows.
3.10.7 Circulatine Water (N71)
<LATER>
3.11 Station Auxiliary Systems 3.11.1 Makeup Water (P10)
<lATER>
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3.11.3 Turbine Buildine Cooline Water fPfl
<LATER>
3.11.4 Chilled Water System (P251 (LATER>
3.11.5 Condensate Storage and Transfer (P30)
(LATER>
3.11.6 Oxyzen Iniection (P32)
<LATER>
3.11.7 Process Sampling (PS3)
<LATER>
3.11.8 Reactor Service Water (P41)
(LA*IT.R>
3.11.9 Turbine Building Service Water (P42)
(LATER>
3.11.10 Service Air and Instrument Air iP51. P521
<LATER>
3.11.11 Mich Pressure Nitrocen Gas Supply (P54)
(LATER>
3.11.12 Auxiliary Boiler (P62)
The Auxiliary Boiler shall provide non-nuclear auxiliary steam for the following functions:
IIcating of water via heat exchangers for various building heating.
a.
h Main Turbine gland :,ealing during plant startup if main stearn is not available ind after I
reactor isolation from power operation.
Warming of offgas preheater.
c.
d.
Preoperational testing of OfTgas Sprem equipment and steamjet air ejectors.
Anti-freeze control of makeup water to the condensate storage tank. and of water in the c.
condensate storage tank ifit is located outdoors.
f.
Chemical cleanup (flushing and cleaning systems after maintenance and prior to system initial startup).
g.
Evaporation ofliquid nitrogen for inerting the Containment.
Auxiliary steam to the Radioactive Waste Management Systems is not required. The steam requirements for each of the above functions is provided in the appended data sheet referenced in paragraph 1.3.3.a Nto ac7 mEv 4;se) 4
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<LATER>
3.11.14 Post Accident Samplieg (P911
<LATER>
3.11.15 Iron injection (P951
<LATER>
3.12 Station Electrical Swtems 3.12.1 Electrical Power Distributis.m (RIO) 3.12.1.1 On-site electric power sha!! be supplied from one of three sources: plant main generator, utility power grid, or aflktite power source. Plant loads shall be supplied from the main plant generator through the unit auxiliary transformer.e during normal plant operation.
The utility power grid shall be the normal preferred power supply when the turbine generator is tripped. Off-site power shall be the alternate preferred power supply.
3.12.1.2 A generator breaker shall be utilized in the power supply system as a standard feature to allow the unit auxiliary transformers to remain connected to the grid and to supply loads by backfeeding from the switchyard following a turbine trip.
3.12.2 Unit Auxiliary Transformers (Rll). The set of the Unit Auxiliary Transformers feeding the on-site power distribution systems shall be composed of two three phase transformers with an installed spare. Each transformer shall be sized to carry 50% of the auxiliary power required for full power operation of the plant or 100% of the auxiliary power required for operation of the plant at 67% power, which ever is the largest.
3.12.3 Isolated Phase Bus (R131. The Isolated Phase Bus duct system shall provide cl5ctrical interconnections from the Main Generator output terminals to the low voltage generator breaker and from the low voltage generator breaker to the low voltage terminals of the main transformer, and the Unit Auxiliary Transformers. During the time the main generator is oft line the low voltage generator breaker shall be open and power fed to the unit auxiliary transformers by back feeding from the main transformer. During startup the generator breaker shall be closed at about 7% power to provide power to the main and the unit auxiliary transformers for normal operation of the plant.
3.12.4 Non-Secrerated Phase Bus (R21). 'Ibe Non-Segregated Phase Bus shall provide the electrical interconnection between the Uni Auxiliary Transformers and their associated 6.9 kV metal-clad switchgear.
3.12.5 Metal Clad Switchcear (RTdl. The Metal Clad Switchgear shall distribute 6.9 kV power.
Circuit breakers shall be drawout type, stored energy vacuum breakers. The switchgear interrupting rating shall be determined in accordance with requirements of ANSI C37.10.
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33 nsv A 3.12.6 Power Centers (R23) 3.12.6.1 Power for 480 V auxiliaries shall be supplied from load centers consisting of 6.9 kV/480 V transformers and associated metal clad switchgear. Power transformers shall be ventilated dry-type,3 phase, connected delta-primary and wye-secondary. Primary voltage rating shall be 6.9 kV. Secondary rating shall be 480 Y/277 V. Each secondary neutral shall be solidly grounded. He variety of transformer sizes shall be kept to a minimum.
Transformer impedance may be adjusted to aid in the selection of economical switchgear ratings.
3.12.6.2 480 V switchgear shall be free-standing indoor metal enclosed gear with manual tie breakers. All circuit breakers shall be of the large stored energy air circuit breaker type. Solid state tripping shall be used. The switchgear short circuit ratings shall be determined in accordance with requirements of ANSI C37.13. Ilowever,if molded case breakers are used, the short circuit ratings shall be based on NEMA AIM.
3.12.7 Motor Control Centers (R24) 3.12.7.1 480 V Motor Control Centers shall prcwide power to motors 90 kW (121 hp) and smaller, control power transformers, process heaters, motor-operated valves and other small electrically operated auxiliaries, including 480-120 V and 480-240 V transformers. Starters for the control of 460 V motors 90 kW (121 hp) and less shall be motor control center mounted across-the-line, magnetically operated, air break circuit breaker type. They shall be the combination type embodying a circuit breaker of 25,000 A (symrnetrical) interrupting capacity to provide instantaneous only short circuit protection and disconnecting means, and a magnetic contactor to provide overload and under voltage protection.
3.12.7.2 Four Class 1E 480 V motor control centers shall be provided to serve as points of isolation between the Non-Class IE power centers and the Class lE 480 V loads. Each motor control center shall supply power to loads of one of the four divisions only. The power supply for these motor control centers shall be Non Class IE from the power centers which receive power from the standby din,el generators.The motor control centers for divisions I and 3 shall be supplied from one diesel generator. Divisions 2 and 4 motor control centers shall be supplied from the other diesel generator.
3.12.7.3 Combination starters for 460 V motors shall be mounted in the factory. assembled motor control centers whenever feasible. Rese motor control centers shall bc NEMA type B, class 11 construction located in switchgear rooms or other suitable non hazardous areas.
Maximum allowabic short circuit capacity at a motor control center bus shall bc 22.000 A (symmetrical). Where starters are not located on motor control centers, they shall have individual enclosures to conform with the design conditions and area classification at their location. Approximately 15% spare space will be provided in motor control centers. Back-tm back motor control centers may be utilized if required by space considerations.
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GENuc/esEnerF 2SA672S suo. 69 arv A 3.12.8 Raceway (RSI) 3.12.8.1 Class !E circuits shall be routed only through Class IE raceways in accordance with the separation criteria specified in the document referenced in paragraph 1.3.1.1. The need for associated circuits shall be avoided if possible. If associated circuits are required they shall be routed in accc rdance with IEEE-384 and Regulatory Guide 1.75. No other circuits shall be routed in Class 1E raceways. Redundant Non-Class IE circuits shall be run in separate raceways to the extent possible. Raceway and circuit routing shall comply with BTP-CMEB 9.5.1 (SRP 9.5.1), except that separation by distance outside the control room or primary containment is not acceptable in general. Separation between redundant divisions shall be by fire barriers with a minimum rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, if possible. Exceptions to this requirement shall be analyzed andjustified as being acceptable on an individual basis. Safe shutdown of the plant shall be possible with complete burnout of any single fire area without recovery. Separate raceways shall be provided for each category of cables shown below. If possible, the arrangement of vertically stacked trays, top to bottom, shall be:
a.
Medium-voltage power b.
Low-voltage power AC and DC cables Inw-voltage control cables (120 VAC,125 VDC) c.
d.
Instrumentation e.
Lighting 3.12.8.2 All Class IE raceways (tray and conduit) shall have Seismic Category I supports.
Non-Class lE raceways and supports shall be designed for SSE loads where there is a potential hazard from these raceways and supports to a Seismic CategoryI installation. This should be done on a selective basis using a criterion adapted for the purpose.
3.12.8.3 Raceways shall be kept a reasonable distance from heat sources such as steam piping, boilers, high and low pressure heaters, and any other actual or potential heat source.
Generally, it is recommended that trays and exposed conduits be no closer than 30.5 cm (12 in) above the top ofinsulation on hot pipes from 66 C to 121 C (150 F to 250 F) and 45.7 cm (18 in) above the top ofinsulation on hot pipes from 122 C to 260 C (251 F to 500*F) where the trays or conduits are crossing perpendicular to the pipes. When running parallel to the pipes, these distance should be increased by 15.2 cm (6 in). All cases of heat source crossing should be carefully enluated and supplemental heat shielding used, if necessary.
3.12.8.4 Class IE and Non-Class IE raceways may be supported by the same seismic category I support (hanger), pro ided that the seismic category I support is capable of supporting the potential additional load imposed by failure of any related non-category I support.
3.12.8.5 Raceway system design shall comply with IEEE-422, cable system in power generation station.
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,_ 20 ruv A 3.12.9 Cab!c (RSS) 3.12.9.1 A.; a minimum, the cables used in the plant shall be rated for a maximum conductor temperature of 90*C (194*F) under emergency overload conditions and 250*C (482*F) under short circuit condition. The conductor shall be son drawn copper with tin or lead alloy coating. Cable construction shall be in accordance with ICEA and/or IEC standards. All cables shall be capable of passing the IEEE-383 flame test. The cable shall be classified as being used for the following applications:
hiedium-voltage power,8 kV shielded a.
h Low-voltage power, 600 V c.
Control,600 V d.
Instrumentation & special purpose e.
Lighting, 600 V 3.12.9.2 Wire and cable for service at 6.9 kV shall be single or triplexed, shielded, cross-linked-polyethylene (XLP) type conductor insulation, flame-retardant Neoprene, flame-retardant Ilypalon, or flame-rctardant cross-linked-polyethlene jacket, rated 8.0 kV, and may be run in tray or conduit. Wire for application at 600 V and below shall be sing!c or multi-conductor, Dame retardant cross-linked polyethylene type conductor insulation, with a flame-retardant Neoprene, name retardant Hypalon, or flame retardant cross-linked-polyethylene jacket. Insulation rating shall be 600 V. hiulti-conductor cable shall be used wherever feasible.
3.12.9.3 Instrumentation cables shall have cross-linked polyethylene (XLP) insulaticm with Hypalon, neoprene or equaljacket. Special purpose cables shall have insulation andjacket material in accordance with the application.
3.12.9.4 Lighting wire shall have XLP insulation or equal.
3.12.9.5 Tefzel (Registered Trademark of E.1. Du Pont de Nemours & Company) insulated cable types, or other approved types, where available is an acceptable alternate. Tefrel insulated cable sizes are not to be used for sizing raceways.
3.12.10 Plant Grounding (R34). Plant Grounding shall be designed using IEEE-80,IEEE-142, and IEEE837 standards as guides and to the requirements of document DC El, Electrical Design Criteria.
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3.12.11 Electric Penetrations (RSM 3.12.11.1 Power, control and instrument circuits shall pass through the containment building wall in electrical penetration assemblics. Separate penetrations shall be provided for medium-voltage, low-voltage power, lighting, control, and instrument circuits. Containment electrical penetration shall comply with US Regulatory Guide 1.63 and IEEE 317.
3.12.11.2 Class-1E circuit separation groups designated Division 1,2,3,4, and Non Class 1E circuits shall be rtm through separate penetration assemblics. These penetrations shall be h>cated so that the phpical separation as specified in the document referenced in paragraph 1.3.1.1 shall be maintained between separation groups.
3.12.11.3 Containment electrical penetrations shall be rated and protected so that a failure of any circuit of a penetration shall not result in exceeding the maximum current versus time capability of the penetration in the event of a single failure of a protective device.
3.12.11.4 All medium and low voltage power circuits passing through electrical penetrations shall be provided with primary and backup protective devices.
3.12.11.5 The design of the control circuits, control power circuits, and instrumentation circuits passing through electrical penetrations shall be based on minimizing the need to protect the penetration from the effects of fault or overload currents.
3.12.11.6 Where protecdve devices are used to protect the penetrations, the penetrations shall be designed to withstand the maximum possible fault and overload currents for the time suflicient for operation of the backup protecthc devices in case of failure of the primary protective devices.
3.12.11.7 All Class 1E equipment shall be environmentally and seismically qualified to assure the execution ofits safety function.
3.12.13 Standbv AC Power Supply (R401 3.12.13.1 Two standby non+afety-related AC diesel power supplies shall proside separate sources of onsite power for various auxiliary and investment protecdon load groups when the nonnal and alternate preferred power supplies are not available. The diesel generators shall be configured to provide power to the permanent non-safety-related buses.
3.12.13.2 Transfer to the standby diesel generators shall be automatic when all other power supplies capable of feeding the bus are not available. Transfer back to the preferred power supply shall be a manual operation.
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,_..n mv A 3.12.14 Direct Current Power Supolv (R42) 3.12.14.1 The plant direct current (DC) power supply system shall consist of two non-divisional 250VDC power supplies, two non-divisional 125 VDC power supplies, and four divisional Class IE 125 VDC power supplies.
3.12.14.2 The DC electrical power sources shall have sufIicient battery capacity to power all safety-related systems requiring electrical power concurrently and to attain prompt shutdown and continue maintenance of the plant in a safe condition under all postulated accident conditions. No active power transfers shall be required for DC safety-related supplies in the event of a main generator trip.
3.12.14.3 The safety-related batteries shall be sized to support the essentialloads for a period of time consistent with station blackout requirements (para. 2.5).
3.12.15 Vital (Uninterruptible) AC Power Supply (R46) 3.12.15.1 Four independent Class IE 120 VAC vital instrument supplies shall be prodded to supply four divisions of safety-related instrumentation and controls. Each dtal instrument l
supply shall consist of an inverter (supplied from Clau IE DC bus), distribution bus, backup (vohage regulating) transformer and a bypass switch. Under normal operation, the inverter shall supply the vital AC bus from the DC source, Each unit shall consist of an automatic (static) transfer switch to an AC source to ensure continuous power to the load in case ofloss of the DC source. A manual bypass switch shall be provided for periodic testing and maintenance without power interruption to the load. The system shall be 120 VAC, single j
phase,60 hz and solidly grounded and designed to maintain steady state voltage to within 12% of120 V and frequency to within 1/2 hz of 60 hz.
3.12.15.2 Two Non-Class IE 120 VAC vital instrument supplies shall be provided to supply power to instrumentation and control of non-safety-related systems. Each non-safety-related vital instrument supply configuration is the same as the Class IE vital AC power supply above except, the inverter shall be supplied from a Non-ClasvlE DC bus and the backup transformer shall be fed from a Non-Class IE 480 V MCC.
3.12.15.3 Two Non Class lE 120 VAC vital power supplies for the plant computers shall bc provided. The irwerters shall be fed from the 250 VDC buses and the backup transformer shall be fed from Non-Class IE 480 VAC MCCs.
3.12.16 Instrument and Control Power Supp1v (R47L Indhidual voltage regulating transformers shall supply 120 VAC instrument power. There shall be four safety-related and two non-safetyrelated 120VAC instrument power supplies. Each Class IE divisional transformer shall be supplied from a 480 V MCC which supplies safety-related loads of one division only. Each Non-Class IE transformer shall be supplied from a non divisional 480 V NEO 907 ( Arv 4/RA)
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3.12.17 Communication (R51). The SBWR plant shall have a Communication System consisting of a dial telephone sptem (by Applicant), a power actuated paging facility and a sound powered telephone sprem for maintenance and repair.
3.12.18 Lichtine and Serviciy Power Supply (R52). The Ughting and Servicing Power System provides adequate illumination during both normal and accident operating conditions. The plant lighting is divided into four sul> systems:
Normal (non-safety-related) lighting (AC) a.
h Standby (safety-related) lighting (AC) c.
Standby emergency (DC safety related) lighting (DC) d.
Normal emergency (DC non safety-related) lighting (DC)
The Lighting Sptem shall be designed to comply with the electrical standards of NFPA and OSHA as applicable to for safety of personnel, plant and equipment. The lighting guidelines' e
shall be based on !!!uminating Engineering Society (lES).
3.13 Power Transmission Systems (S-series) i 3.13.1 It is an interface requirement for the Applicant to provide information on the power transmission system. The interface point between the SBWR design and the utility design for the main generator output is at the connection of the isolated phase bus to the main power transformer low voltage terminals. The rated conditions for this interface are 600 MVA and 25 kV. It is a requirement that the utility provide sufficient impedance in the main power transformer and the high voltage circuit to limit the primary side maximum available fault current contribution from the system to no more than 275 kA symmetrical and 340 kA asymmetrical at 5 cycles from inception of a three phax fault at the primary terminals of the transformer. When all equipment and sptem parameters are known, a refined calculation based on the known values with a fault located at the generator side of the generator breaker shall be made. This may allow a lower impedance for the main power transformer, if desired.
3.13.2 The second power interface occurs at the high voltage terminals of the reserve auxiliary transformer, The rated load is 7.5 MVA at a 0.9 power factor. The voltage and frequency shall be the utilitics standard with the actual values to be determined as a part of the detailed plant design.
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,4 nEv A 3.13.3 Protective relaying interfaces for the two power 1 stem interfaces shall be dcEned as a part of the detailed plant design.
3.14 Gontainment and Emironmental Control Systems 3.14.1 Containment Sntem (TIO) 3.14.1.1 The Containment System shall consist of a Containment Vessel (CV) with the following components:
A suppression pool shall be provided to absorb energy by condensing steam during SRV a.
discharges RPY depressurization or in-containment pipe breaks. The pool shall be sized such that the pool temperature does not exceed 71.l*C (160*F) for design basis events.
The pressure suppression flow path shall be designed to entrain radioactive materials by routing flow through the pool during and following a LOCA.
h A drywell region shall be provided as a leak-tight gas space surrounding the reactor vessel to contain radioactive fitssion products, steam and water prior to directing them to the suppression pool, c.
A wetwell region shall be provided as a gas space sized to collect and retain the drywell air following a pipe break in the drywell without exceeding the CV design pressure, d.
A drywell/,wetwell vent system shal! be provided to channel blowdown flow from the drywell to discharge into the suppression poolin a horitantal direction below the water surface. The vent submergence shall be selected to ensure steam condensation for all dmign basis conditions, including suppression pool drawdown after a LOCA.
Wetwell/drywell vacuum breakers shall be provided to prevent overloading on.the c.
containment linens during depressurization of the drywell under <later> conditions.
3.14.2 Containment Vessel and Structures (Til. T12) 3.14.2.1 The Containment Vessel shall be a seismic CategoryI, structure. It shall be a cylindrical vessel and shall directly support the upper pools. The walls of the upper pools shall be integrated with the top slab of the Containment Vessel to provide structural capability for LOCA and testing pressures.
3.14.2.2 It shall be possible to test the containment integrity and leak tightness at period in t ervals.
3.14.2.3 The Containment Vessel and the reactor building shall share the basemat.
3.14.2.4 The Containment Vessel shall accommodate the severe accident load combination given in the document referenced in paragraph 1.3.1.n.
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75 nev A 3.14.2.5 The Containment Vessel shall be capable of being flooded up to an elevation <Iater>
above that corresponding to the top of the reactor core, if necessary, for long-term recovery following an accident, concurrent with presst.re of (Later> within the remaining Containment Vessel gas space.
3.14.3 Passive Containment Cooling System (T15) 3.14.3.1 The Passive Containment Cooling System (PCCS) shall consist of 2 (or 3) low pressure PCCS loops. Each PCCS loop consists of a condenser which is open to the primary containtnent, a drain line to the GDCS (gravity driven cooling system) pool, a vent discharge line which is submerged in the pressure suppression pool and a cooling pool (IC/PCCS pool).
The PCCS condenxrs share a cooling pool with the ICs of the Isolation Condenser System (B32).
3.14.3.2 The PCCS loops receive a steamtas mixture supply directly from the drywell. The PCCS loops are driven by the pressure difference created between the containment drywell and the suppression pool during a LOCA so require no sensing, control, logic or power-actuated devices for its functioning. The PCCS loops are an extension of the safety-related containment and do not have isolation valves. Provisions are included in the PCCS to conduct Type B containment leakage tests separately from Type A containment leakage tests (Reference 10CFR50, AppendixJ).
3.14.3.3 Two (or three) PCCS condensers shall be sized to maintain the Containment within its pressure limits for design basis accidents, including severe accidents (reference para. 2.4).
3.14.3.4 The IC/PCCS cooling pool is sized to remove post-LOCA decay heat for a minimum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and with cooling pool inventory not being replenished.
3.14.4 Atmosoheric Control (T31) 3.14.4.1 The Atmospheric Control System (ACS) shall include means to remove radiolytic or metal & water reactor hydrogen that will be relcased to the containment during a LOCA.
)
3.14.4.2 The ACS shall include a non-safety-related containment inerting system to handle hydrogen that might be generated during a severe LOCA or accident.
3.14.5 Dnwell Cooling (T4D 3.14.5.1 Non-safety-related cooling shall be achieved using coolers supplied by the reactor component cooling water system (RCCWS).
3.14.6 Flammability Control (T491
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vs ncv A 3.14.7 Suppression Pool Temt>erature Monitorine (T53)
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l 3.15 Structures and Servicine Systems 3.15.1 Iurbine Pedestal (U24)
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3.15.2 Cranes. Hoists and Elevators (U311
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3.15.3 Turbine Building IWAC (U39). The Turbine Buildmg HVAC shall have the following features:
provide fresh air and keep the turbine building at slightly negative pressure a.
b.
regulate the building temperature remove smoke and noxious fumes from the turbine building c.
d.
control the spread of radioacthity within the turbine building failure of the turbine building HVAC shall not cause an off-site dose to exceed the e.
10CFR100 limit.
3.15.4 Reactor Building HVAC (UiO). The Reactor Building HVAC consists of three scperate sub-systems. The first sub-system shall maintain the MCR & TSC Environs. The second sub-system shall control the balance of the reactor building. The third sul> system passively maintains habitable conditions in the operating areas in an emergency.
3.15.4.1 The MCR & TSC HVAC normal operation HVAC sul> system has the follo$ing featurcs:
Provide fresh air and keep the MCR & TSC at slightly positive pressure.
a.
b.
Regulate the rooms temperature and humidity.
4 c.
Remove smoke and noxious fumes from the MCR & TSC.
d.
Isolate on emergency.
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77 arv A 3.15.4.2 The balance of the reactor building INAC sub-sptem shall have the following features:
Provide fresh air and keep the reactor building at slightly negative pressure.
a.
h Regulate the building temperature, Remove smoke and noxious fumes from the reactor building.
c.
d.
Isolate on emergency to provide a holdup for any radioactive gasses.
Control the spread of radioactivity within the reactor building.
c.
f.
Failure of the reactor building IWAC shall not cause an off-site dose to exceed the 10CFR100 limit.
3.15.4.3 The Control Room Emergency Habitability sul> system shall maintain habitable conditions in the emergency operating areas, in accordance with the requirements specified-in paragraph 2.16.
3.15.5 Other Building HVAC (U41)
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t 3.15.6 Fire Protection (U43) 3.15.4.1 Fire barrier, smoke control, and fixed fire suppression sptems shall be installed throughout the plant to satisfy the design bases requirements. Standpipe and hose systems, together with portable extinguishers shall be provided in all building areas. The type of fire suppression shall be determined based on combustibles present and the safety function of the area and to provide protection against exposure fires. A comprehensive fire detection, alarm, and control system shall be utilized throughout the plant. The primary operation of this i
system is automatic, govemed by local control panels throughout the plant. A main fire and detector status panelin the Main Control Room will be provided to monitor and receive alarm and trouble signals from local panels.
3.15.4.2 A single active failure or a crack in a moderate <nergy line (pipe) in a fire suppression system should not impair both the primary and backup fire suppression capability. For example, neither the failure of a fire pump, its power supply or controls, nor a crack in a moderate-energy line in the fire suppression system, should result in loss of function of both sprinkler and hose standpipe systems in an area protected by such primary and backup systems.
3.15.4.3 As a minimum, the fire suppression system should be capable of delivering water to manual hose stations located within hose reach of areas containing equipment required for safe plant shutdown following a SSE. To achieve this, one source of water supply (storage source, pump, water main, and standpipes) to areas containing safety-related equipment shall NC O 907 (PCV 4t$e) l
GENudewEnergy og,s,g,
,8 REV A be analyzed to withstand the effects of SSE and remain ftmctional. Therefore, at least one means of fire fighting will always be available.
3.15.4.4 All safety-related systems are provided with redundant equipment. The Fire Protection System shall be designed so that the most serious consequence of a fire is that it t
may incapacitate only one safety-related division, leaving the redundant division, or divisions, to perform its safety function.
3.15.4.5 The design of the fire barrier system shall be such that complete burn out of any single fire arca without recovery will not prevent safe shutdown of the plant. It shall be assumed that the temperature profile of the fire area experiencing the fire follows the ASTM E-Il9 time-temperature curve for a duration of three hours.
3.15.4'.6 Allowable combustible loadings consistent with the capabilities of the fire confinement (barrier), smoke control, and fire suppression systems shall be established.
i 3.15.4.7 Fire Protection shall meet the requirements of BTPCMEU 9.5.1 (SRP 9.5.1).
3.15.7 Buildines & Services (U65 to U73. U42. U44. U45. U50)
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3.16 Intake Structure and Servicing Enuinment (W-wries)
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3.17 Yard Structures and Eouipment (Y-series)
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