ML20045E710

Forwards Submittal 3 of Sys 80+ Design Descriptions & Associated ITAAC for Review & Approval

ML20045E710
Person / Time
Site:	05200002
Issue date:	06/18/1993
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LD-93-096, LD-93-96, NUDOCS 9307020344
Download: ML20045E710 (100)

Text

_. - . _ . . .

A BB ED M ED D ASEA BROWN BOVERi June 18, 1993 LD-93-096 Docket No.52-002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk l~ Washington, DC 20555

Subject:

System 80+" Submittal #3 Design Descriptions and'ITAAC

Dear Sirs:

Enclosed is Submittal #3 of the System 80+ Design Descriptions and associated ITAAC (Inspections, Tests, Analyses and Acceptance Criteria) which are submitted for review and approval. l 1

ABB-CE is performing an Integrated Review of the CESSAR-DC and Design Descriptions /ITAAC to ensure consistency among and within these documents. It is _ possible that changes to the enclosed material may be necessary should the review uncover any l l inconsistencies. It is our intention to incorporate such changes I in our final amendment targeted for June 30, 1993. j i

In addition, resolution of the pressurized post-accident sampling issue could cause revision of the enclosed Design Descriptions and ITAAC for the Process Sampling System.

Please feel free to query us as the staff evaluates this submittal.

You may contact me or Mr. John Rec at (203) 285-2861 for assistance in this matter.

Very truly yours, W

C. B. Brinkman Acting Director Nuclear Systems Licensing gdh/lw

Enclosure:

As Stated cc: T. Boyce (NRC)

T. Wambach (NRC)

P. Lang (DOE) r J. Trotter (EPRI)

A. Heymer (NUMARC)

ABB Combustion Ergineering Nuclear Power )0 b

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com.sm,rmneew m mo %se m exf ueme na ree '9" Pod CWe Ban MO Fan f?03y 285 9512 9307020344 930618 PDR w,-w c yec2 mou ne nm covato wson A ADDCK 05200002 pyg _ _ _ _ _ _ _ _ _ _ _ - - -

h Sh f*LBHlh ASEA BROWN BOVERI D

June 18, 1993 LD-93-096 Docket No.52-002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

System 80+" Submittal #3 Design Descriptions and ITAAC

Dear Sirs:

Enclosed is Submittal #3 of the System 80+ Design Descriptions and associated ITAAC (Inspections, Tests, Analyses and Acceptance Criteria) which are submitted for review and approval.

ABB-CE is performing an Integrated Review of the CESSAR-DC and Design Descriptions /ITAAC to ensure consistency among and within these documents. It is possible that changes to the enclosed material may be necessary should the review uncover any inconsistencies. It is our intention to incorporate such changes in our final amendment targeted for June 30, 1993.

In addition, resolution of the pressurized post-accident sampling issue could cause revision of the enclosed Design Descriptions and ITIAC for the Process Sampling System.

Please feel free to query us as the staff evaluates this submittal.

You may contact me or Mr. John Rec at (203) 285-2861 for assistance in this matter.

Very truly yours, C. B. Brinkman Acting Director Nuclear Systems Licensing gdh/lw

Enclosure:

As Stated cc: T. Boyce (NRC)

T. Wambach (NRC)

P. Lang (DOE)

J. Trotter (EPRI)

A. Heymer (NUMARC)

ABB Combustion Engineering Nuclear Power Ccerbustion 1.ngmevg hc 1D30 P'oued K0 Road Teetyene (203! 6881911 Post O't.ce Bo= 500 Faa (203) 285 9512 husor, Connectcst D6095 0500 Teen 99297 COMBEN WSOR

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7. En A Netsm 2_ TorK O A.11yd e O J. Lu o s n . kAtenui2v 0 A Ke sh e n G.NeGS C- File Special Comments Nr,tc A: This form must be included with the attached document when designated as a quality record.

Form 8 0090244 (Rev.1*B1)

O4 h

SYSTEM 80+"

1.1 DEFINITIONS The following definitions apply to terms used in the Design Descriptions and associated  ;

ITAAC:

Acceptance Criteria means the performance, physical condition, or analysis result for a structure, system, or component that demonstrates the Design Commitment is met.

Analysis means a calculation, mathematical computation, or engineering or technical evaluation. Engineering or technical evaluations could include, but are not limited to, <

comparisons with operating experience or design of similar structures, systems, or components.

As-built means the physical properties of a :;tructure, system, or component following the completion of its installation or construction activities at its final location at the plant site.

Basic Configuration (for a Building) means the arrangement of structural features (e.g.,

floors, ceilings, walls, columns, and doorways) and any structures, systems or components which are specified in the building Design Description.

Basic Configuration (for a System) means the functional arrangement of structures, systems, or components specified in the Design Description and the verifications for that system specified in Section 1.2.

Chnnnelmeans either an arrangement of components and their interconnects that generates a signal to initiate an action or to indicate a condition, or a distinct path among multiple paths within instrumentation and control equipment. ]

Design Commitment means that portion of the Design Description that is verified by ITAAC.

Design Description means that portion of the design that is certified.

Division (for electrical systems or equipment)is the designation applied to a given safety-related system or set of components which are physically, electrically, and functionally independent from other redundant sets of components.

Division (for mechanical systems or equipment)is the designation applied to a specific set of safety-related components within a system.

l 1.1 06-18-93 1

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SYSTEM 80+"

Inspect or Inspection mean visual observations, physical exaninations, or reviews of records based on visual observation or physical examination that compare the structure, system, or component condition to one or more Design Commitments. Examples include walkdowns, configuration checks, measurements of dimensions, or non-destructive examinations.

Test means the actuation, operation, or establishment of specified conditions to evaluate the performance or integrity of as-built structures, systems, or components, unless explicitly stated otherwise.

Type Test means a test on one or more sample components of the same type and manufacturer to qualify other components of that same type and manufacturer. A Type Test is not a test of the as-built structures, systems, or components.

1.1 06-18-93

q ,

SY!TmM 80+" l 1

1 L2 GENERAL PROVISIONS The following general provisions are applicable to the Design Descriptions and associated ITAAC.

Verifications For Basic Configuration For Systems i Verifications for Basic Configuration of systems include and are limited to inspection of the system functional arrangement and the following inspections, tests, and analyses:

(1) Inspections, including non-destructive examination (NDE), of the as-built, pressure boundary welds for ASME Code Class 1,2, or 3 components identified in the Design Description to demonstrate that the requirements of ASME Code Section III for the ,

quality of pressuie boundary welds are met. i (2) Tests, or tests and analyses, of the Seismic Category I mechanical and electrical equipment (including connected instrumentation and controls) identified in the Design Description, including associated anchorage, to demonstrate that the equipment is qualified to withstand design basis dynamic loads without loss of its safety function. >

(3) Tests, or tests and analyses, of the Class 1E electrical equipment identified in the Design Description (or on accompanying Figures) to demonstrate that it is qualified to withstand the emironmental conditions that would exist during and following a design basis accident without loss of its safety function for the time needed to be functional. These emironmental conditions, as applicable to the bounding design basis accident (s), are as follows: expected time-dependent temperature and pressure profiles, humidity, chemical efrects, radiation, aging, submergence, and their synergistic ,

effects which have a significant effect on equipment performance. As used in this paragraph, the term

• Class 1E electrical equipment" constitutes the equipment itself, connected instrumentation and controls, connected electrical components (such as cabling, wiring, and terminations), and the lubricants necessary to support performance of the safety functions of the Class 1E electrical components identified ,

in the Design Description, to the extent such equipment is not located in a mild environment during or following a design basis accident.

4 Electrical equipment environmental qualification shall be demonstrated through -i analysis of the emironmental conditions that would exist in the location of the equipment during and following a design basis accident and through a determination that the equipment is qualified to withstand those conditions for the time needed to be functional. This determination may be demonstrated by:

f (a) Testing of an identical item of equipment under identical or similar conditions with a supporting analysis to show that the equipment is qualified; or 1.2 06-18-93

SYSTEM 80+"

(b) testing of a similar item of equipment under identical or similar conditions with a supporting analysis to show that the equipment is qualified; or (c) experience with identical or similar equipment under identical or similar conditions with supporting analysis to show that the equipment is qualified; or (d) analysis in combination with partial type test data that supports the analytical assumptions and conclusions to show that the equipment is qualified. l (4) Tests or type tests of active safety-related motor-Operated Valves (MOVs) identified in the Design Description to demonstrate that the MOVs are qualified to perform their safety functions under design basis differential pressure, system pressure, fluid temperature, ambient temperature, minimum voltage, and minimum and/or maximum stroke times.

Treatment of Individual Items The absence of any discussion or depiction of an item in the Design Description or ,

accompanying Figures shall not be construed as prohibiting a licensee from utilizing such an item, unless it would prevent an item from performing its safety functions as discussed or depicted in the Design Description or accompanying Figures. l When the term " operate," " operates," or " operation"is used with respect to an item discussed in the Acceptance Criteria,it refers to the actuation and running of the item. When the term

" exist," " exists," or " existence" is used with respect to an item discussed in the Acceptance Criteria, it means that the item is present and meets the Design Description.

Implementation of ITAAC The ITAAC are provided in tables with the following three-column format:

Inspections Desien Commitment Tests Analyses Acceptance Criteria Each Design Commitment in the left-hand column of the ITAAC tables has an associated Inspections, Tests, or Analyses (ITA) requirement specified in the middle column of the tables.

The identification of a separate ITA entry for each Design Commitment shall not be construed to require that separate inspections, tests, or analyses must be performed for each Design Commitment. Instead, the activities associated with more than one ITA entry may be combined, and a single inspection, test, or analysis may be sufficient to implement more than one ITA entry.

1.2 06-18-93

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l

SYSTEM 80+"

An ITA may be performed by the licensee of the plant, or by its authorized vendors, contractors, or consultants. Furthermore, an ITA may be performed by more than a single i individual or group, may be implemented through discrete activities separated by time, and may be performed at any time prior to fuel load (including before issuance of the Combined Operating License for those ITAAC that do not necessarily pertain to as-installed equipment). Additionally, an ITA may be performed as part of the activities that are required to be performed under 10 CFR Part 50 (including, for example, the Quality Assurance (QA) program required under Appendix B to Part 50); therefore, an ITA need not be performed as a separate or discrete activity.

Discussion of Matters Related to Operations In some cases, the Design Descriptions in this document refer to matters that relate to operation, such as normal valve or breaker alignment during normal operation modes. Such discussions are provided solely to place the Design Description provisiou m context (e.g., to explain automatic features for opening or closing valves or breakers upon off-normal  ;

conditions). Such discussions shall not be construed as requiring operators during operation to take any particular action (e.g., to maintain valves or breakers in a particular position during normal operation). ,

Interpretation of Figures In many but not all cases, the Design Descriptions in Section 2 include one or more Figures. '

The Figures may represent a functional diagram, general structural representation, or other general illustration. For I&C systems, Figures also represent aspects of the relevant logic of the system or part of the system. Unless specified explicitly, the Figures are not indicative of the scale, location, dimensions, shape, or spatial relationships of as-built structures, systems, and components. In particular, the as-built attributes of structures, systems, and components may vary from the attributes depicted on the Figures, provided that those safety functions discussed in the Design Description pertaining to the Figure are not adversely affected.

1 i

i t

I.2 06-18-93 ,

I

1.4 FIGURE LEGEND AND ABBREVIATION LIST (The figure legend and abbreviation list are provided for information only)

FIGURE LEGEND instrumentation Toxic Gas Detector b Flow Instrument b Temperature instrument b Radiation Instrument @

Differential Pressure Instrument Pressure Instrument g Level Instrument g Current Instrument g Humidity Detector g Ultrasonic Instrument g Smoke Detector g Sensor Annunciator (Alarm) $  ;

1 Annunciator Symbols For:  !

High High HH High H

)

Low L l Low Low LL 1.4 06-18-93

_ _ _ _ - - . _ _ _ _ _ _ _ _ _ _ _ _ +

FIGURE LEGEND (continued)

Valves Gate Valve Cl><3 Globe Valve [>Stl3 Check Valve Butterfly Valve l%l Ball Valve @

Relief Valve Three Way Valve Valve Type Not Specified c4 Valve Ooerators Operator Of Unspecified Type Fluid Powered Operator Motor Operator Solenoid Operator m

Diahragm Operator Hydraulic Operator -

Pnuematic Operator Position Indications For Hydraulic And Pneumatic Ooerators

-Fails As is FAI

-Falls Closed FC

-Falls Open F0 Mechanical Eouloment Positive Displacement Pump _

k -

06-18-93

i FIGURE LEGEND (continued) l Centrifugal Pump +0"  :

l l

Pump Type Not Specified + i Header [ ]  !

m l l

Tank V

Filter F OR FILTER 1

l S

Strainer Flexible Connection Delay Coil M Orifice !l 1

Venturi l Compressor Or Fan i[O Air Distribution Device  !

Air Distribution Header lill Vaneaxial Fan M

+ +

Heat Exchanger V h Vacuum Breaker Vent o v I4 06-18-93

1 FIGURE LEGEND (cong;p.edl Damoers T '

Manually Operated Damper f OR s s l

Remotely Operated Damper  ; i Louver Fire Damper kr  :

3. '

Smoke Damper k

S j

Back Draft Damper Pumo Drivers .

Turbine Drive Motor Drive Electrical Eauioment Battery g Circuit Breaker A Disconnect Link n 1A 06-18-93 l

l

FIGURE LEGEND (continued)

Multiplexer E lsolation b Transformer Miscellaneous A System Or Component 1 -----~l That is not Part Of The l l Defined System i______

Containment Containment with Penetration - , , , - , ,

1 - - , , ,

T Building Separation iuuuuunuun ASME Code Class Break j An ASME Code class _ break is identified by a single horizontal or vertical line perpendicular to the designated location for the class break, as shown in the example below. l l

1 l ASVE CODE SEl~T ON Ill CLASS l (NOTE 1) 1_ 11 l i

X +

N l Notes:

1. The header, "ASME Code Section 111 Class", must appear at least once on each figure on which ASME class breaks are shown, but need not appear at every class break shown on a figure.

E Indicates Non-ASME Code Section ill 1.4 06-18-93 I

SYSTEM 80+"

ABBREVIATION LIST Abbreviation Meaning AAC Alternate AC Source A/C Air Conditioning ADM Atmospheric Dump Valve AFAS Alternate Feedwater Actuation Signal ALMS Acoustic Leak Monitoring System APC Auxiliary Process Cabinet APS Alternate Protection System AVS Annulus Ventilation System BAC Boric Acid Concentrator CCCT Containment Cooler Condensate Tank CCS Component Control System CCVS Control Complex Ventilation System CCW Component Cooling Water CCWLLSTAS Component Cooling Water Low Level Surge Tank Actuation CCWS Component Cooling Water System CEA Control Element Assembly CEACP CEA Change Platform CEAE CEA Elevator CEDMCS Control Element Drive Mechanism Control System CEDM Control Element Drive Mechanism CET Core Exit Thermocouple CFR Code of Federal Regulations CFS Cavity Flooding System CGCS Combustible Gas Control System CGS Compressed Gas Systems CH Channel ,

CIAS Containment Isolation Actuation Signal 1.4 06-18-93

SYSTEM 80+"

- ABBREVIATION LIST (Continued)

Abbreviation Meaning CIS Containment Isolation System CIV Containment Isolation Valve COL Combined Operating License CONT Containment CPC Core Protection Calculator CPVS Containment Purge Ventilation System CRS Control Room Supervisor CSAS Containment Spray Actuation Signal CSB Core Support Barrel CSS Containment Spray System CST Chemical Sample Tank CT Combustion Turbine / Generator CVAP Comprehensive Vibration Assessment Program CVCS Chemical and. Volume Control System CWT Chemical Waste Tank DBVS Diesel Building Ventilation System DEMIN Demineralized DFSS Diesel Fuel Storage Structure DIAS Discrete Indication and Alarm System DIAS-N Discrete Indication and Alarm System - Channel N DIAS-P Discrete Indication and Alarm System - Channel P DNBR Departure From Nucleate Boiling Ratio

/ DPS Data Processing System DSW Dry Solid Waste DVI Direct Vessel Injection DWMS Demineralized Water Makeup System ECWS Essential Chilled Water System 1.4 06-18-93

(

SYSTFM,.,ny $

p .. T ABBJg ?;d.iiQi _ LIST (Continued) ijdq,[

Meaning Abbreviat( (

EDG Emergency Diesel Generator EDT Equipment Drain Tank EFAS Emergency Feedwater Actuation Signal EFDS Equipment and Floor Drainage System EFW Emergency Feedwater EFWS Emergency Feedwater System EFWST Emergency Feedwater Storage Tank ENS Emergency Notification System EPDS Electrical Power Distribution System ESF Engineered Safety Features ESFAS Engineered Safety Features Actuation System ESF-CCS Engineered Safety Features - Component Control System EWT Equipment Waste Tank FBOC Fuel Building Overhead Crane FBVS Fuel Building Ventilation System FDT Floor Drain Tank FHS Fuel Handling System FTC Fuel Temperature Coefficient FTS Fuel Transfer System GCB Generator Circuit Breaker GWMS Gaseous Waste Management System HA High Activity HDR Header HFE Human Factors Engineering IUTC Heated Junction Thermocouple HPN Health Physics Network HSI Human-System / Interface

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SYSTEM 80+"

ABBREVIATION LIST (Continued)

Abbreviation Meaning HVAC Heating, Ventilating, Air Conditioning HVT Holdup Volume Tank HX Heat Exchanger l HZ Hertz IAS Instrument Air System ICI In-Core Instrument ILRT Integrated Leak Rate Test INIT Initiation INJ Injection INST Instrumentation IPSO Integrated Process Status Overview IRWST In-containment Refueling Water Storage Tank ITP Interface and Test Processor IVMS Internals Vibration Monitoring System IWSS In-containment Water Storage System IX Ion Exchanger IA Low Activity LBB Leak-Before-Break LHST Laundry & Hot Shower Tank LOCA Loss-of-coolant Accident LOOP Loss-of-Offsite-Power LPMS Loose Parts Monitoring System LS Liquid Sample LTOP Low Temperature Overrpressure Protection LWMS Liquid Waste Management System MCC Motor Control Center MCR Main Control Room 1.4 06-18 93

)

1 SYSTEM 80+"

ABBREVIATl MSI (Continued) g l

l Abbreviation Meaning MCRACS Main Control Room Air Conditioning System  !

1 MDNBR Minimum Departure From Nucleate Boiling Ratio l

MFIV Main Feedwater Isolation Valve i MG Main Generator l

MOV Motor Operated Valve MPC Moderator Pressure Coefficient MSIS Main Steam Isolation Signal )

MSIV Main Steam Isolation Valve MSLB Main Steam Une Break MSSS Main Steam Supply System MSSV Main Stran Safety Valve MSVH Main Steam Valve House j MTC Moderator Temperature Coefficient NA Nuclear Annex l

NAVS Nuclear Annex Ventilation System )

i NCW Normal Chilled Water NCWS Normal Chilled Water System NDE Non-destructive Examination l NFE New Fuel Elevator i

NFS Nuclear Fuel System J NI Nuclear Instrumentation NI Structures Nuclear Island Structures NIMS NSSS Integrity Monitoring System NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission PA Public Address PABX Private Automatic Business Exchange 1.4 06-18 93

i SYSTEM 80+"

ABBREVIATION LIST (Continued) .

Abbreviation Meaning PAMI Post Accident Monitoring Instrumentation P-CCS Process-Component Control System l PCPS Pool Cooling and Purification System

\

PCS Power Control System PCS/P-PCCS Power Control System / Process-Component Control System  :

PERMSS Processing and Effluent Radiological Monitoring and i Sampling System PPC Plant Protection Calculator PPS Plant Protection System l PRA Probability Risk Assessment j PSS Process Sampling System PSWS Potable and Sanitary Water Systems )

PZR Pressurizer RAT Reserve Auxilliary Transformer l

RB Reactor Building i 1

RCGVS Reactor Coolant Gas Vent System j RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RDS Rapid Depressurization System RDT Reactor Drain Tank RM Refueling Machine RPS. Reactor Protective System RSP- Remote Shutdown Panel RSR ' Remote Shutdown Room RSSH Resin Sluce Slurry Header l l

RT Reactor Trip  !

1,4 06-18 93  !

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1 SYSTEM 80+" '

ABBREVIATION LIST (Continued)

Abbreviation Meaning RTSG Reactor Trip Switchgear RV Reactor Vessel i RWBVS Radwaste Building Ventilation System SAFDL Specified Acceptable Fuel Design Limit SB Shield Building SBVS Subsphere Building Ventilation System SCS Shutdown Cooling System SDS Safety Depressurization System SFHM Spent Fuel Handling Machine SFP Spent Fuel Pool SFPCS Spent Fuel Pool Cooling System SG Steam Generator SGBS Steam Generator Blowdown System SGDT Steam Generator Drain Tank SI Safety Injection SIAS Safety Injection Actuation Signal SIS Safety Injection System SIT Safety Injection Tank SSE Safe Shutdown Earthquake SSW Station Service Water SSWS Station Service Water System  ;

l SWMS Solid Waste Management System l TBCWS Turbine Building Cooling Water System l TBSWS Turbine Building Service Water System TBV Turbine Bypass Valve TC Thermocouple TGSS Turbine Gland Sealing System 1.4 06-18-93

Ii SYSTEM 80+

ABBREVIATION LIST (Continued)

Abbreviation Meanine TSC Technical Support Center TSCACS Technical Support Center Air Conditioning System UGS Upper Guide Structure -

UHS Ultimate Heat Sink UAT Unit Auxillary Transformer  ;

1 UMT Unit Main Transformer VCT Volume Control Tank VDU Video Display Unit WMT Waste Monitor Tank  !

l WSW Wet Solid Waste l l

i l

l I

I

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SYSTEM 80+"

2.2.1 NUCLEAR FUEL SYSTEM Design Description The Nuclear Fuel System (NFS) generates heat by a controlled nuclear reaction and transfers the heat generated to the reactor coolant. The NFS consish of an arrangement in the reactor vessel of fuel assemblies and control element aanblics

! (CEAs). He NFS has the safety-related functions of providing a barrier against the release of radioactive material generated by nuclear reactions in the nuclear fuel and providing a means to make the reactor core suberitical.

l The Basic Configuration of the fuel assembly, the CEAs and their arrangement in the l reactor core is as shown on Figures 2.2.1-12.2.1-2, and 2.2.1-3. The reactor core has a maximum of 241 fuel assemblies and a minimum of 93 CEAs. j Each fuel assembly has fuel rods, spacer grids, guide tubes, and upper and lower end fittings. In each fuel assembly, a maximum of 236 locations are occupied by fuel rods.

l The remaining 20 locations are subdivided into five symmetric regions, each of which contains a guide tube that provides a channel for insertion of a CEA finger or an in-core instrument. Each guide tube is attached to fuel assembly spacer grids and to fuel assembly upper and lower end fittings to provide a structural frame to position the I fuel rods.

l Each CEA has a maximum of 12 CEA fingers, each containing neutron absorbing material within a cylindrical, sealed metal tube. The CEA fingers are held in position at one end and are spaced to allow entry into the guide tubes of fuel assemblies.

Each fuel rod has fissile material in the form of ceramic pellets. T!.: fuel pellets in l cach fuel rod are contained within a cylindrical, scaled metal tube. Fuel rods can also contain neutron absorbing material.

A fuel assembly can also have rods containing neutron absorbing material that l displace fuel rods. One or more fuel assemblies can have a neutron generating source j located within a guide tube.

l The stresses of the fuel assembly and fuel assembly components (excluding the fuel rods) are predicted to not exceed the specified limits for normal operation and the specified limits for design basis events.

The stresses and strains of the fuel rod cladding are predicted to not exceed the specified limits for normal operation.

The cumulative fatigue damage factors of the fuel assembly, fuel assembly components I and fuel rod cladding under cyclic loading conditions are predicted to not exceed I specified limits.

I i

2.2.1 06-17-93 l l

i SYSTEM 80+"

Based on consideration of potential collapse of the fuel rod cladding, fuel rod  !

mechanical integrity is predicted to be maintained during the design lifetime of the l fuel rod. l Based on consideration of potentially excessive cladding pressure loading, fuel rod mechanical integrity is predicted to be maintained during the design lifetime of the  !

fuel rod. ,

ne CEAs are capable ofinsertion into the core during all modes of plant operation within the insertion time limits assumed in the plant safety analyses.

The CEA cladding stresses and strains are predicted to not exceed the specified limits ,

for normal operation. He CEA cladding stresses are predicted to not exceed the 6 specified limits for design basis events.

In the power operating range, the combined response of the fuel temperature coefficient (FTC), the moderator temperature coefficient (hiTC), the moderator void coefficient (hfVC) and the moderator pressure coefficient (hiPC) to an increase in -

reactor thermal power is predicted to be a decrease in reactivity.

The potential amount and rate of reactivity insertion from the reactivity control systems under normal operation and postulated reactivity accidents are predicted to not result in (i) violation of the specified acceptable fuel design limits (SAFDLs) for any single failure of the reactivity control systems (excluding CEA ejection), (ii) J damage to the reactor coolant pressure boundary (RCPB), or (iii) disruption of the  !

reactor core or reactor internals which would impair the ability to provide safety j injection of reactor coolant.

The power distribution and power peaking in the nuclear fuel are predicted to not l result in violation of the SAFDLs for normal operation and anticipated operational occurrences.

The amount of negative reactivity available from insertion of withdrawn CEAs is predicted to meet the specified excess CEA worth requirement for power operating  ;

conditions.

During normal operation and anticipated operational occurrences, the departure from nucleate boiling ratio (DNBR) on any fuel rod is predicted to be greater than or equal to the minimum departure from nucleate boiling ratio (hfDNBR).

1 During normal operation and anticipated operational occurrences, the peak temperature of the nuclear fuel in any fuel rod is predicted to be less than the fuel melt temperature.

1 2.2., 06-17-93 1 i

SYSTEM 80+"

The measured primary coolant flow rate with the four reactor coolant pumps in operation is greater than or equal to a specified minimum and less than or equal to a specified maximum reactor coolant system (RCS) flow rate.

Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Nuclear Fuel System.

2.2.1 06-17-93

SYSTEM 80+"

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p . : qF44E . q pay i s H s

- L

+-

h ,

] 12 ELEMENT CEAS H 4EtEuEnrccAS DENOTES SPARE CEA LOCATIONS

_S t

FIGURE 2.2.1-3 NUCLEAR FUEL SYSTEM ARRANGEMENT l

l

j?YSTEM 80+= TAHLE 2.2.1-1 NUCLEAR FUEL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 1.a) 'Ihe Basic Configuration of the fuel 1.a) Inspection of the fuel assembly and 1.a) For the components and equipment assemblies and the CEAs is as shown on CEA configurations will be conducted. shown on Figures 2.2.1-1 and 2.2.1-2, Figures 2.2.1-1 and 2.2.1-2. the fuel assemblies and the CEAs conform with the Basic Configuration.

1.b) The Basic Configuration of the Nuclear 1.b) Inspection of the Initial Test Program 1.b) The Initial Test Program (Section 2.11)

Fuel System arrangement is as shown on (Section 2.11) requirements for has a requirement to verify that the fuel Figure 2.2.1-3. verification of the fuel assembly and assembly and CEA arrangements con-CEA arrangements will be conducted. form to Figure 2.2.1-3.

2.a) The stresses of the fuel assembly and 2.a) Analyses of fuel assembly and fuel 2.a) The analyses conclude that fuel assembly components (exchiding the assembly component (excluding the fuel fuel rods) are predicted to not exceed rod) stresses will be performed for (i) the general primary membrane stress in the specified limits for normal normal operation. the fuel assembly and fuel assembly operation. components (excluding the fuel rods) during normal operation will not exceed the material design stress intensity value, and (ii) the sum of the general primary membrane stress and the primary bending stress in the fuel assembly and fuel assembly components (excluding the fuel rods) during normal operation will not exceed the product of the material design stress intensity value and the component shape factor.

2.2.1 06-17-93 i

_ _ . - _ _ _ - - _ . - - - _ _ . _ _ _ - - _ _ _ - _ _ _ . _ . _ - - - - _ _ _ _ _ - - - _ . - - -- - , . . -n - --, . - - , - - - . . . - - - - - . , , - -

SYSTEM 80+ TABLE 2.2.1-1 (Continued)

NUCLEAR FUEL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria J)esign Commitment Inspections. Tests. Analyses Acceptance Criteria 2.b) The stresses of the fuel assembly and 2.b) Analyses of fuel assembly and fuel 2.b) The analyses conclude that the fuel fuel assembly components (excluding the assembly component (excluding the fuel assembly and fuel assembly component fuel rods) are predicted to not exceed rod) stresses will be performed for (excluding the fuel rod) stresses will not the specified limits for design basis design basis events. exceed the specified limits for design events. basis events.

3. The stresses and strains of the fuel rod 3. Analyses of fuel rod cladding stresses 3. The analyses conclude that cladding are predicted to tet exceed the and strains will be performed for normal specified limits for normal operation. operation. (i) the maximum pnmary tensile ctress in the fuel rod cladding during normal operation will not exceed two-thirds of the minimum unitradiated yield strength of the material at the applicable temperature, and (ii) the net unrecoverable circumferential strain in the fuel rod cladding during normal operation will not exceed 1 %.

4. The cumulative fatigue damage factors 4. Analyses of fuel assembly, fuel 4. The analyses conclude that the fuel of the fuel assembly, fuel assembly assembly component and fuel rod assembly, fuel assembly component and components and fuel rod cladding under cladding cumulative fatigue damage fuel rod cladding cumulative fatigue cyclic loading conditions are predicted factors will be performed for cyclic damage factors during cyclic loading to not exceed the specified limits, loading conditions. conditions will not exceed 0.8.

2.2.1 06-17-93

SYSTEM 80+- TABLE 2.2.1-1 (Continued) .

NUCLEAR FUEL SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria i

Desian Commitment inspections. Tests. Analyses Acceptance Cdteria

5. Based on consideration of potential 5. Analyses will be performed for fuel rod 5. The analyses conclude that, based on collapse of the fuel rod cladding, fuel mechanical integrity based on consideration of potential collapse of the rod mechanical integrity is predicted to consideration of potential collapse of the ~

fuel rod cladding, fuel rod mechanical  ;

be maintained during the design lifetime fuel rod cladding. integrity will be maintained during the of the fuel rod. design lifetime of the fuel rod.

6. Based on consideration of potentially 6. Analyses will be performed for fuel rod 6. "Ihe analyses conclude that, based on excessive cladding pressure loading, fuel mechanical integrity based on consideration of potentially excessive 4 rod mechanical integrity is predicted to consideration of potentially excessive cladding pressure loading, fuel rod be maintamed during the design lifetime cladding pressure loading. mechanical integrity will be maintained of the fuel rod, during the design lifetime of the fuel rod.

7. The Control Element Assemblies 7. Tests and analyses of CEA insertion 7. The Initial Test Program (Section 2.11)

(CEAs) are capable of insertion into the times will be performed as part of the specifies a test to demonstrate that CEA core during all modes of plant oreration Initial Test Program (Section 2.1I). insertion times are less than or equal to within the insertion time limits assumed specified insertion time limits.

in the plant analyses.

t i

9 2.2.1 06-17-93 1____._. ..._...a__._____.._m_ ..__m___._m. ._____ _ _ _ . _ _ _ _ _ _ _ _ _ . . - . . _ . - , - - , -

_m .<._--,,,%.. iv., . , . , , , . . . ~ . , . , , * , . . .rm,, . ,.s. .-uw-.., .- , , -..,, , ,

SYSTEM 80+ TABLE 2.2.1-1 (Continued)

NUCLEAR FUEL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 8.a) The CEA cladding stresses and strains 8.a) Analyses of CEA cladding stresses and 8.a) He analyses conclude that are predicted to not exceed the specified strains will be performed for normal limits for normal operation. operation. (i) the general primary membrane stress in the CEA cladding during normal operation will not exceed the material design stress intensity value, (ii) the sum of the general primary membrane stress and the calculated primary bending stress in the CEA cladding during normal operation will not exceed the product of the material design stress intensity value and the component shape factor, and (iii) the net unrecoverable circumferential strain in the CEA cladding during normal operation will not exceed I %.

8.b) The CEA cladding stresses are predicted 8.b) Analyses of CEA cladding stnas will 8.b) The analyses conclude that the CEA to not exceed specified limits for design be performed for design basis events. cladding stresses will not exceed the basis events. specified limits for design basis events.

2.2.1 06-17-93

SYSTEM 80+ TABLE 2.2.1-1 (Continucd)

NUCLEAR FUEL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

9. In the power operating range, the 9. Tests and analyses will be performed as 9. He Initial Test Program (Section 2.11) combined response of the fuel part of the Initial Test Program (Section specifies tests and analyses to temperature coefficient (FTC), the 2.11) to determine the combined demonstrate that, in the power operating moderator tetuperature coefficient response of the FTC, the MTC, the range, the combined response of the (MTC), the moderator void coefficient MVC and the MPC to an increase in FFC, the MTC, the MVC and the MPC (MVC) and the moderator pressure reactor thermal power. to an increase in reactor thermal per coefficient (MPC) to an increase in will be a decrease in reactivity.

reactor thermal power is predicted to be a decrease in reactivity.

10. He potential amount and rate of 10. Analyses will be performed to determine 10. The analyses conclude that the potential reactivity insertion from the reactivity the potential amount and rate of amount and rate of reactivity insertion control systems under normal operation reactivity insertion from the reactivity fmm the reactivity control systems lead and postulated reactivity accidents are control systems under normal operation to normal operation and postulated predicted to not result in (i) violation of and postulated reactivity accidents. reactivity accident sequences _which (i) the SAFDLs for any single failure of the will not violate the SAFDLs for any reactivity control systems (excluding single failure of the reactivity control CEA ejection), (ii) damage to the systems (excluding CEA ejection), (ii)

RCPB, or (iii) disruption of the core or will not damage the RCPB, and (iii) will other reactor intemals which impairs the not impair the effectiveness of safety effectiveness of safety injection of injection of reactor coolant.

reactor coolant.

I1. He power distribution and power 11. Analyses will be performed to determine 11. The analyses conclude that the MDNBR peaking in the nuclear fuel are predicted the minimum departure from nucleate and PLIIGR will not exceed the to not result in violation of the SAFDLs boiling ratio (MDNBR) and peak linear SAFDLS during normal operation and for normal operation and anticipated heat generation rate (PLHGR) during anticipated operational occurrences.

operational occurrences. normal operation and anticipated operational occurrences.

2.2.1 06-17-93

SYSTEM 80+ TABLE 2.2.1-1 (Continued)

NUCLEAR FUEL SYSTEM InsDections. Tests. Analyses. and Acceptance Criteria Desirn Commitment Inspections. Tests. Analyses Acceptance Criteria

12. He amount of negative reactivity 12. Analyses will be performed to determine 12. The analyses conclude that the amount available from insertion of withdrawn the amount of reactivity available from of negative reactivity available from CEAs is predicted to meet the specified insertion of withdrawn CEAs for power insertion of withdrawn CEAs is greater excess CEA worth requirement for operating conditions. than or equal to the specified excess power operating conditions. CEA worth requirement for power operation conditions.

13. During normal operation and anticipated 13. Analyses will be performed to determine 13. The analyses conclude that, during operational occurrences, the departure the DNBR during normal operation and normal operation and anticipated from nucleate boiling ratio (DNBR) is anticipated operational occurrences. operation occurrences, the DNBR on predicted to be greater than or equal to any fuel rod will be greater than or the minimum departure from nucleate equal to the MDNBR.

boiling ratio (MDNBR) on any fuel rod.

14. During normal operation and anticipated 14. Analyses will be performed to determine 14. The analyses conclude that during operational occurrences, the peak the peak temperature of the nuclear fuel normal operation and anticipated temperature of the nuclear fuel in any in a fuel rod during normal operation operational occurrences, the peak fuel rod is predicted to be less than the and anticipated oprational occurrences. temperature of the nuclear fuel in a fuel fuel melt temperature, rod will be less than the fuel melt temperature.

15. The measured primary coolant flow rate 15. Tests and analyses will be performed as 15. The Initial Test Program (Section 2.11) with the four reactor coolant pumps in part of the Initial Test Program (Section specifies a test to determine that the operation is greater than or equal to a _ 2.11) to determine the primary coolant primary coolant flow rate is within a specified minimum and less than or flow rate with the four reactor coolant specified range.

equal to a specified maximum RCS flow pumps in operation.

rate.

2.2.1 06-17-93

SYSTEM 80+"

233 REACTOR COOLANT SYSTEM COMPONENT SUPPORTS Design Description ne reactor vessel, the steam generators, the reactor coolant pumps and the pressurizer are supported by the reactor coolant system (RCS) component supports.

The RCS component supports permit movement of the RCS components due to  ;

expansion and contraction of the RCS. The component supports are safety related.

I The RCS component supports are located within the centainment. l l

The four reactor vessel support columns vertically support the reactor vessel and l accommodate horizontal thermal expansion. Each reactor vessel nozzle cold leg l forging mates with a reactor vessel support column and serves as a key which mates ,

with a keyway. Lower keys protruding from the reactor vessel mate with a slot in each support column base plate. The slot in the support column base plate serves as a keyway. These horizontal keys and keyways guide the vessel during expansion and contraction of the RCS, maintain the vessel centerline position, and laterally support the vessel.

Each steam generator (SG) is supported at the bottom by an integral skirt attached to a sliding base plate resting on bearings. The bearings allow the SG to move as the RCS expands and contracts. Keys and keyways within the sliding base guide the movement of the SG during expansion and contraction of the RCS and limit movement of the SG bottom in the direction at right angles to the direction of motion during RCS expansion and contraction. The upper portion of the SG is supported by a system of keys, keyways and snubbers. The upper SG support system guides the top of the steam generator during expansion and contraction of the RCS and laterally supports the SG.

Each reactor coolant pump (RCP) is supported by vertical columns, lower and upper horizontal columns, and snubbers. The columns provide vertical and horizontal support of the RCP, while allowing movement of the RCP during expansion and contraction of the RCS.

He pressurizer is supported at the bottom by an integral skirt. Keys and keyways provide lateral support of the upper portion of the pressurizer. .

The reactor vessel columns, the SG sliding base, bearings, and snubber assembly, and the RCP vertical columns, horizontal columns and snubbers are ASME Code Section III Class 1 and are qualified Seismic Category I.

2.3.3 06-18-93

SYSTEM 80+"

Inspection, Test, Analyses, and Acceptance Criteria Table 233-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Reactor Coolant System Component Supports.

i 1

L 9

4

{

06-18 93  ;

1

SYSTEM 80+= TABLE 233-1 REACI'OR COOLANT SYSTEM COMPONENT SUPPORTS Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The RCS component supports permit 1. A test of the RCS will be performed to 1. Required gaps exist for the RCS movement of the RCS components due monitor thermal motion during heatup component supports.

to expansion and contraction of the and cooldown of the RCS.

RCS.

2. The reactor vessel columns, SG sliding 2. Inspection will be performed for the 2. ASME Code Section III Design Reports base, bearmgs, and snubber assembly, existence of the ASME Code Section III exist for:

and the RCP vertical columns, Design Reports for the following horizontal columns, and snubbers are supports:

ASME Code Section III Class 1.

reactor vessel columns, reactor vessel columns, SG sliding base, bearings, and snubber SG sliding base, bearings, and snubber assembly, and assembly, and RCP vertical columns, horizontal RCP vertical columns, borizontal columns and snubbers. columns and snubbers.

1 i

l 2.3.3 06-18-93

l SYSTEM 80+"

2.3.4 NSSS INTEGRITY MONITORING SYSTEM Design Description The NSSS Integrity Monitoring System (NIMS) is a non-safety-related instrumentation and control system which consists of the Internals Vibration Monitoring System (IVMS), the Acoustic Leak Monitoring System (ALMS), and the Loose Parts Monitoring System (LPMS). 'Ihe NIMS provides data to the data processing system (DPS). The IVMS provides data from which changes in the motion of the reactor internals can be detected. The ALMS provides data in response to i high acoustic levels originating from a reactor coolant pressure boundary (RCPB) leak. The LPMS provides data in response to vibration of the RCPB associated with loose parts within the RCPB. -

The NIMS is located in the nuclear island structures.  :

Displays of the NIMS instrumentation exist in the main control room (MCR) or can be retrieved there. ,

Inspections, Tests, Analyses and Acceptance Criteria .

, Table 23.4-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the NSSS Integrity Monitoring System.

1 t

l 2.3.4 06-17-93 i

l

SYSTEM 80+= TABLE 23.4-1 NSSS INTEGRITY MONITORING SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. He IVMS provides data from which 1. Tests willbe performed on the IVMS by 1. The IVMS provides data to the DPS in changes in the motion of the reactor providing a test signal simulating a time- response to the test signal, intemals can be detected. varying signal from the ex-core neutron detector channels.

2. The ALMS provides data in response to 2.a) Inspection of the as-built ALMS 2.a) ALMS sensors are provided in locations high acoustic levels originating from a configuration will be performed. specified in Table 2.3.4-2.

RCPB leak.

2.b) Tests will be performed on the ALMS 2.b) The ALMS provides data to the DPS in by providing a test signal simulating response to the test signal.

high acoustic levels.

3. He LPMS provides data in response to 3.a) Inspection of the as-built LPMS 3.a) LPMS sensors are provided in locations vibration of the RCPB associated with configuration will be performed. specified in Table 2.3.4-3.

loose parts within the RCPB.

3.b) Tests will be performed on the LPMS 3.b) The LPMS provides data to the DPS in by providing a test signal simulating response to the test signal.

motion of the RCPB locations.

4. Displays of the NIMS instrumentation 4. Inspection for the existence or 4. Displays of the NIMS instrumentation exist in the MCR or can be retrieved retrievability in the. MCR of exist in the MCR or can be retrieved there, instrumentation displays will be there.

! performed.

I i

2.3.4-1 06-17-93

SYS'mM 80+" l 1

TABLE 23.4-2 SENSOR LOCATIONS FOR ACOUSTIC LEAK MONTIURING SYSTEM .

I COMPONENT NUMBER OF LOCATION  !

SENSORS Reactor Coolant Pump 4 (1 per pump) Seal i Steam Generators 2 (1 per SG) Frimary side, manway Hot Legs 2 (1 per Leg) Reactor vessel outlet nozzle l Cold Legs 4 (1 per Leg) Reactor vessel inlet nozzle l Reactor Vessel 3 Upper head, CEDM nozzles Reactor Vessel 1 Lower head, instrument nozzle Pressurizer Safety Valves 4 (1 per valve) Discharge line ]

I Pressurizer 1 Heater region 2.3.4 06-17-93 4

SYSTEM 80+"

TABLE 23.4-3 SENSOR LOCATIONS FOR LOOSE PARTS MONITORING SYSTEM COMPONENT NUMBER OF SENSORS LOCATION Reactor Vessel 3 Lower Head  ;

3 Upper Head Steam Generator 1 4 Primary (inlet plenum)

Primary (outlet plenum)

Secondary (economizer region)

Secondary (can deck region) i f

Steam Generator 2 4 Primary (inlet plenum)

Primary (outlet plenum)

Secondary (economizer 3 region)

Secondary (can deck region) i a

t i

e 2.3.4 06-17-93 ,

SYSTEM 80+"

2.43 COMBUSTIBLE GAS CONTROL SYSTEM '

Design Description The Combustible Gas Control System (CGCS) is used to maintain hydrogen gas concentration in Containment at a level which precludes an uncontrolled hydrogen and oxygen recombination within Containment following design basis and beyond design basis accidents.

The CGCS has hydrogen gas analyzers and hydrogen mitigation devices which can include hydrogen recombiner units, hydrogen ignitors, and self-powered (catalytic) hydrogen recombination devices. Hydrogen recombiner units,if provided, are safety-related.

A minimum of two hydrogen analyzers are provided to measure hydrogen gas concentration in Containment.

Hydrogen mitigation devices are installed in Containment, or if not installed in

• Containment, can be connected by piping to Containment.

Electrically powered hydrogen mitigations devices receive power from their respective Class 1E Division.

Displays of the CGCS hydrogen analyzer instrumentation exist in the main control room (MCR) or can be retrieved there.

Controls exist in the MCR to energize and de-energize the hydrogen analyzers and the electrically powered hydrogen mitigation devices in Containment.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.4.3-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Combustible Gas Control System.

2.4.3 06-17-93

SYSTEM 80+ TAILLE 2.4.3 COMBUSTIHLE GAS CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. A minimum of two hydrogen analyzers 1. Inspection of the as-installed hydrogen 1. At least two hydrogen gas analyzers are are provided to measure hydrogen gas analyzers will be performed. Shop tests installed which can measure hydrogen concentration in Containment. will be performed on the hydrogen gas concentration in Containment.

analyzers for calibration.

2. Ilydrogen mitigation devices are in- 2. Inspection of the hydrogen mitigation 2. IIydmgen mitigation devices are in-stalled in Contaimnent, or if not devices will be performed. stalled in Containment, or if not in-installed in Containment, can be stalled in Containment, can be connected connected by piping to Containment. by piping to Containment.

3. Electrically powered hydrogen miti- 3. Tests will be perfonned on the elec- 3. 'Ihe test signal exists only in the elec-gations devices receive power from thJr trically powered hydrogen mitigation trically powered hydrogen mitigation respective Class IE Division. devices by providing a test signal in devices under test.

only one Class 1E Division at a time.

4.a) Displays of the CGCS hydrogen 4.a) Inspection for the existence or re- 4.a) Displays of the CGla Sydrogen analyzer instrumentation exist in the trieveability in the MCR of instru- analyzer instrumentation exist in the MCR or can be retrieved there. mentation displays will be performed. MCR or can be retrieved there.

4.b) Controls exist in the MCR to energize 4.b) Tests will be performed using the CGCS 4.b) CGCS controls in the MCR operate to and de-energize the hydrogen analyzers controls in the MCR. energize and de-energize the hydrogen and the electrically powered hydrogen analyzers and the electrically powered mitigation devices in Containment. hydrogen mitigation devices in Con-tainment.

2A.3 06-17-93

i SYSTEM 80+"

2.4.7 IN-CONTAINMENT WATER STORAGE SYSTEM Design Description The In-containment Water Storage System (IWSS) includes the in-containment refueling water storage tank (IRWST), the holdup volume tank (HVT), and the cavity flooding system (CFS).

The IRWST provides horated water for the safety injection system (SIS) and the containment spray system (CSS). It is the primary heat sink for discharges from the reactor coolant system (RCS) pressurizer safety valves and the safety depressurization system (SDS) rapid depressurization subsystem. It is the source of water for the CFS.

It is the source of water to fill the refueling pool via the SIS and CSS. The IRWST and IRWST instrumentation are safety-related.

The HVT collects water released in Containment during design basis events and returns water to the IRWST through spillways. It also collects component leakage not routed to other drain systems inside Containment and receives water discharged from the IRWST by the CFS.

The CFS is used to provide water to flood the reactor cavity in response to beyond design basis events.

The IWSS is located in the Containment.

The Basic Configuration of the IWSS is as shown on Figure 2.4.7-1.

The IRWST has a volume above the SIS / CSS pump suction line penetrations to permit proper SIS and CSS operation following design basis events. The IRWST has a total volume that permits dilution of radionuclides from core and RCS release following design basis loss-of-coolant accidents (LOCAs).

Stainless steel baskets containing disodium phosphate are located in the HVT.

The ASME Code Section III Class for the IWSS pressure retaining components is as shown on Figure 2.4.7-1.

The safety related equipment shown on Figure 2.4.7-1 is qualified Seismic Category I.

Displays of IWSS instrumentation shown on Figure 2.4.7-1 exist in the main control room (MCR) or can be retrieved there.

Controls exist in the MCR to open and close those power operated valves shown on Figure 2.4.7-1.

2.4.7 06-17-93

3 i

SYSTEM 80+"

IWSS alarms shown on Figure 2.4.7-1 are provided in the MCR.

The power operated valves and IRWST instrumentation shown on Figure 2.4.7-1 are powered from their respective Class 1E Division. Within the CFS, each of the four valves in the spillways from the IRWST to the HVT is powered from a different Class 1E bus. 'l Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.7-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Incontainment Water Storage System.

i l

I i

i 9

2.4.7 06-17-93

SYSTEM 80 +* ED Ei]

SDS SDS ,p;' .g' ,

I VB VB .-- ~N N 1r U N N ses _h" L* _sc3 N g 2 g g l

SIS- - > -

sis- - >

IN CONTAINMENT g' - - WPSS REFUELING WATER STORAGE HOLDUP TANK VOLUME - - > PSS

• g

>< =

• R

>< = E

• H

• M _ ___

i I

[><_

= >< = i REACTOR i

• M

• M I CAVITY I INSIDE CONTAINMENT g J . .'. l. - . .

)( y

( m I

e I

i OUTSIDE I I I I CONTAINMENT y yyy SIS SIS SIS SIS NOTES:

1. THE IRWST AND IRWST INSTRUMENTATION SHOWN ARE SAFETY-RELATED

2. THE POWER OPERATED VALVES AND !RWST INSTRUMENTATION SHOWN ARE POWERED FROM THEIR RESPECTIVE CLASS 1E DIVISION

3. * : EQUIPMENT FOR WHICH PARAGRAPH NUMBER 3 OF THE " VERIFICATION FOR BASIC CONFIGURATION FOR SYSTEMS" SECTION OFTHE GENERAL PROVISIONS (SECTION 1.2) APPLIES.

FIGURE 2.4.7-1 IN-CONTAINMENT WATER STORAGE SYSTEM

SYSTEM 80+~ TAHLE 2.4.7-1 IN-CONTAINMENT WATER STORAGE SYSTEM InSricctionS. TcSts. AnalVSes, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the IWSS is 1. Inspection of the as-built IWSS con- 1. For the components and equipment as shown on Figure 2.4.7-1. figuration will be conducted. shown on Figure 2.4.7-1, the as-built IWSS conforms with the Basic Configuration.

2.a) ne IRWST has a volume above the 2.a) Inspection of construction records for 2.a) The IRWST has a useable volume of at

. SIS / CSS pump suction line penetrations the IRWST will be performed. least 495,000 gallons above the SIS / CSS to permit proper SIS and CSS operation pump suction line penetrations.

following design basis events.

2.b) The IRWST has a total volume that 2.b) Inspection of construction records for 2.b) The IRWST has a minimum total permits dilution of radionuclides from the IRWST will be performed. volume of at least 545,800 gallons.

core and RCS release following design basis LOCAs.

3. Stainless steel baskets containing 3. Inspection of the as-built HVT will be 3. Stainless steel baskets containing disodium phosphate are located in the perfonned. disodium phosphate are located in the llVT. HVT.

4. The ASME Code Sectiort Ill IWSS com- 4. A pressure test will be conducted on 4. He results of the pressure test of ponents shown on Figure 2.4.7-1 retain those components of the IWSS required ASME Code Section III portions of the their pressure boundary integrity under to be pressure tested by ASME Code IWSS conform with the pressure testing internal pressures that will be Section Ill. criteria in ASME Code Section 111.

experienced during service.

2.4.7 06-17-93 1

, . _ - . - - - .- . . - - , ,c- +. , - - , ,, , ,,.

SYSTEM 80+" TABLE 2.4.7-1 (Continucd)

IN-CONTAINMENT WATER STORAGE SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 5.a) Displays of the IWSS instrumentation 5.a) Inspection for the existence or re- 5.a) Displays of the instrumentation shown shown on Figure 2.4.7-1 exist in the trievability in the MCR of instru- on Figure 2.4.7-1 exist in the MCR or MCR or can be retrieved there. mentation displays will be performed. can be retrieved there.

5.b) Controls exist in the MCR to open and 5.b) Tests will be performed u6.g J.: !WSS 5.b) IWSS controls in the MCR operate to close those power operated valves controls in the MCR. open and close those power operated shown on Figure 2.4.7-1. valves shown on Figure 2.4.7-1.

5.c) IWSS alarms shown on Figure 2.4.7-1 5.c) Tests of the IWSS alarms shown on 5.c) The IWSS alarms shown on Figure are provided in the MCR. Figure 2.4.7-1 will be performed using 2.4.7-1 actuate in response to signals signals simulating alarm conditions. simulating alarm conditions.

6.a) He power operated valves and IRWST 6.a) Tests will be performed on the IWSS 6.a) A test signal exists only at the IWSS instrumentation shown on Figure 2.4.7-1 components by providing a test signal in components powered from the Class IE are powered from their respective Class only one Class IE Division at a time. Division under test.

IE Division.

6.b) Within the CFS, each of the four valves 6.b) Tests will be performed on the CFS 6.b) A test signal exists only at the CFS in the spillways from the IRWST to the valves by providing a test signal in only valves powered fmm the Class IE bus IIVT is powered from a different Class one Class IE bus at a time. under test.

IE bus.

2.4.7 06-17-93

SYSTEM 80+"

2.5.2 ENGINEERED SAFETY FEATURES - COMPONENT CONTROL SYSTEM Design Description The Engineered Safety Features-Component Control System (ESF-CCS) is a safety-related instrumentation and control system which provides automatic actuation of Engineered Safety Features (ESF) systems upon receipt of ESFinitiation signals from the Plant Protection System (PPS). The ESF CCS also provides the capability for manual actuation of ESF systems, manual control of ESF system components and manual control of other safety-related systems and components identified below.

The ESF-CCS is located in the nuclear island structures.

The Basic Configuration of the ESF-CCS is as shown on Figure 2.5.2-1.

The ESF-CCS is qualified Seismic Category I.

The ESF-CCS uses sensors, transmitters, signal conditioning equipment, and digital equipment which perform the calculations, communications and logic to generate signals to actuate protective system equipment. This equipment is Class 1E.

The ESF-CCS is divided into four divisions. Each division of the ESF-CCS has the following elements, as depicted on Figure 2.5.2-2:

selective 2-out-of-4 logic, component control logic, process instrumentation, signal conditioning equipment, maintenance and test panel, control and display interface devices, and a j master transfer switch. l The four ESF-CCS divisions are physically separated and electrically isolated.

Each ESF-CCS division is powered from its respective Class 1E bus. '

l Each ESF-CCS division receives 4 channels of initiation signals from the PPS which are processed using selective 2-out-of-4 logic to generate actuation signals for the ESF systems controlled by that division. Basic block diagrams for the functional logic used in the ESF-CCS for actuation of ESF systems are shown on Figures 2.5.2-3 and

. 2.5.2-4.

4 2.5.2 06-18-93

4 SYSTEM 80+"  :

I The ESF-CCS provides control capability and, upon receipt ofinitiation signals from the PPS, automatically generates actuation signals to the following ESF systems within allocated response times: ,

safety injection system, t containment isolation system, ,

containment spray system,  ;

main steam isolation, and emergency feedwater system.

The ESF-CCS provides control capability and, upon receipt ofinitiation signals from ,

the PPS, automatically generates actuation signals to the following non-ESF systems:

annulus ventilation system, ,

component cooling water system,  ;

onsite power system, and diesel generators.

The ESF-CCS provides control capability for the following safety-related systems: ,

shutdown cooling system, safety depressurization system, -

atmospheric dump system, i station service water system, heating, ventilating and air conditioning systems, and i bydrogen mitigation devices. ,

Upon receipt of ESF initiation signals for safety injection, containment spray or emergency feedwater, the ESF-CCS initiates an automatic start of the diesel generators and automatic load sequencing of ESF loads. i Upon detecting loss of power to Class 1E Division buses, the ESF-CCS automatically  ;

initiates startup of the diesel generators, shedding of electrical load, transfer of Class -

IE bus connections to the diesel generator, and sequencing of the reloading of safety- .

related loads to the Class IE bus. In performing load sequencing, normally used safety related plant loads are loaded first in a predetermined sequence unless an ESF ,

actuation signal is generated. Upon ESF actuation, the normal load sequence is ,

interrupted and priority is given to loading the actuated ESF systems and associated safety-related systems. The sequence for loading the normally used safety related  ;

plant loads is then resumed. I The ESF-CCS provides interlock control for isolation valves in the shutdown cooling  !

system (SCS) suction lines, the safety injection tank (SIT) discharge lines and the  !

cmergency feedwater (EFW) pump discharge lines. The SCS interlocks prevent the 2.5.2 06-18-93 i

l r

k SYSTEM 80+"  ;

i i

ESF-CCS from generating a signal to open the SCS isolation valves when the RCS pressure is above the entry pressure of the SCS. The SITinterlocks prevent the ESF- i CCS from generating a signal to close the SlTisolation valves when the RCS pressure is above the entry pressure of the SCS. The interlock on the EFW isolation valves j automatically closes the isolation valves on high SG levels when an Emergency  ;

Feedwater Actuation Signalis not present.  ;

The control and display interface devices of the ESF-CCS in the MCR provide for l automatic and manual control of ESF systems and components. In the remote ,

shutdown room, the control and display interface devices provide for manual control' i of ESF system components needed to achieve hot standby. Actuation of master ,

transfer switches it either exit of the MCR transfers control capability from the l control and displa interface devices in the MCR to those in the remote shutdown l room. Each ESF-CCS dhision's maintenance and test panel provides capability to transfer control fro n the MCR to the remote shutdown room for its respective ESF- .

CCS division and to transfer contro! back to the MCR for its respective ESF-CCS ,

division.  !

Diverse manual cctuation switches are provided as an alternate means for manual actuation of ESF components in two divisions of the ESF-CCS as follows:

2 trains of safety injection, 1 train of containment spray,  ;

1 train of emergency feedwater to each steam generator, ,

1 main steam isolation valve in each main steam line, l

1 isolation valve in each containment air purge line, and I letdown isolation valve.

The diverse manual actuation switches provide input signals to the lowest level in the i ESF-CCS digital equipment. Communication of the signals from the switches is diverse from the software used in the higher levels of the ESF-CCS. Actuation of the ,

switches provides a signal which overrides higher level signals, to actuate the l associated ESF component or components. .

Periodic testing to verify operability of the ESF-CCS can be performed with the reactor at power or when shutdown without interfering with the protective function of the system. Capability is provided for testing all functions, from ESF initiating signals received from the PPS through to the actuation of protective system  ;

equipment. The maintenance and test panel provides capability for manual testing l of ESF-CCS functions and hardware. ,

Where the ESF-CCS and the process control system interface with the same l component (e.g., with sensors, signal conditioners, or actuated devices), electrical isolation devices are provided between the process control system and the shared l

2.5.2 06-18-93 4

i

i SYSTEM 80+"  !

component. Electrical isolation devices are provided at ESF-CCS interfaces with the discrete indication and alarm system - channel N (DIAS-N), the data processing system (DPS), the process-component control system (P-CCS), the control and display  ;

interface devices, the master transfer switches and between the signal conditioning  !

equipment and one of the two channels of the discrete indication and alarm system - ,

channel P (DIAS-P), as shown on Figure 2.5.2-2.

ESF-CCS software is designed, tested, installed and maintained using a process which defines the organization, responsibilities, and activities for the software engineering life cycle and which specifies requirements for software quality assurance, verification '

and validation, configuration management, and operations and maintenance, and which incorporates a graded approach according to the software's relative importance to safety. l Setpoints for interlocks and actuation of ESF-CCS safety-related functions are  !

determined, documented, installed and maintained according to methodologies which l specify requirements for documenting the bases for selection of trip setpoints, accounting for instrument uncertainties and drift, testing of instrumentation setpoint i response and replacement of setpoint related instrumentation. The setpoint calculations are consistent with the physical configuration of the instrumentation.  !

l Inspections, Tests, Analyses, and Acceptance Criteria Table 2.5.2-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Engineered Safety Features-Component Control System.

l 1

l

)

l 3

l l

J 2.5.2 4- 06-18-93

- , _ _ ,- _ _ __- . _ _ _ . _ , . _ _ _ _ _ _ , _ . _ . . . ~ __.

1. -r

"~

SYSTEM 80 + l PPSI

. _ __ _ s l

I I ESF-CCS I

I I

I CONTROL & DISPLAY I I INTERFACE DEVICES

! MASTER TRANSFER SWCHES I i COMPONENT I

DIVERSE MANUAL CONTROL I ACTUADON SWirCHES g

J_________q ,

I

,_ DIAS & DPS

_, I y ,

L - _ _ _ _ . .

gi

- i l I- ~PnOCETScIS - - 'l M, ,

m _ _ _ _ _ _ _ .

i~

I POWER CONTROL SYSTEM 'l l SIGNAL I s _ _ __-_ i i I 8

PPS i I l._ _ _ ._ L _ _ _. __ __ _ __ .I 1_ J_I SENSORS I_ JESF_ _ I l _ .,_ ( lCOMPONENTSI FIGURE 2.5.2-1 ENGINEERED SAFETY FEATURES-COMPONENT CONTROL SYSTEM CONFIGURATION

SYSTEM 80 + - -

r tSF,NMA NQ SIGNALS I

, FROM 4 CHANNELS OF PPS ,

y A B C D g ESF-CCS

~~

b

~~-~--

h-l l MAIN CONTROL ROOM DIVISION KEY:

CONTMOL & DISPLAY II II II ly l 4 ,, , g , ,, ,, ,

MAINTENANCE SELECTIVE -" HARD WIRED OR l e & TEST PANEL 24UT-OF-4 DATA UNK g

MASTER TRANSFER SWITCH

, . . , . [,) , , a tooge NON-CONDUCTING ,

' . . . > DATA LINK OR l DtVERSE MANUAL e

, a 1I lr DISCRETE SIGNAL ACTUATION SWITCHES , a"** h (E.G. FIBER OPTIC)

~

""""" b O ISOLATION

?

PHYSICAL SEPARATION l REMOTE SHUTDOWN ROOM _ _ _ *I BETWEEN CHANNELS

! CONTROL & DISPLAY COMPONENT INTERFACE DEVICES 4 ""[4"""""""

• CONTROL L - - ---

_OGIC

~

l TO ON  !

I D3AS - CHANNEL N_

==

_ _ .h . . .......**"

+ -lliu ESF-CCS

-i DIVISION I l= DATA PROCESSING SYSTEM .....h ..

==___= ..I e I_PPS I PROCESS-CCS j JL JL NOTE 1: IMPLEMENTED IN TWO DIVISIONS

• • ""............. SioNAt CONDmONiNo

!_ POWER CONTROL SYSTEM *"

_=__=_ i

.' C Il

!_ DIAS-CHANNEL P _

j -

I i 1r es l ESF SENSORS & I (sJ

, COMPONENTS ,

FIGURE 2.5.2-2 ENGINEERED SAFETY FEATURES-COMPONENT CONTROL SYSTEM ONE DIV!SION AND INTERCONNECTIONS

SYSTEM 80.

ESFAS ESFAS ESFAS ESFAS INITIATION INITIATION INITIATION INITIATION SIGNALS SIGNALS SIGNALS SIGNALS A A A A f A I A f A f \

)flflflf lIlfl fl f l fl Il Il I l fl fl Il f SELECTIVE SELECTIVE SELECTIVE SELECTIVE 2 OUT OF 4 2 OUT OF 4 2 OUT OF 4 2 OUT OF 4 LOGIC LOGIC LOGIC LOGIC lf If II lf COMPONENT COMPONENT COMPONENT COMPONENT CONTROL CONTROL CONTROL CONTROL LOGIC LOGIC LOGIC LOGIC 1 I lf I I lI TRAIN A TRAIN B TRAIN C TRAIN D COMPONENTS COMPONENTS COMPONENTS COMPONENTS FIGURE 2.5.2-3 ESFAS BASIC BLOCK DIAGRAM FOR SAFETY INJECTION ACTUATION AND EMERGENCY FEEDWATER ACTUATION

q t

. SYSTEM 80+* .;

i P

i ESFAS ESFAS INITIATION INITIATION -

SIGNALS SIGNALS -

A A 7

/ T / T j I fl fl fl f l fl fl fI f f SELECTIVE SELECTIVE f 2 OUT OF 4 2 OUT OF 4 i LOGtC LOGIC -l j

i lI l I- .

COMPONENT COMPONENT CONTROL CONTROL LOGIC LOGIC 1 f 1I ,

TRAIN A TRAIN B i COMPONENTS COMPONENTS  ;

i i

?

t

+

FIGURE -2.5.2-4 '

ESFAS BASIC BLOCK DIAGRAM FOR MAIN STEAM.

ISOLATION, CONTAINMENT SPRAY ACTUATION, AND i CONTAINMENT ISOLATION  ;

- .._ ..__. . _ . . . . _ . - . . , _ . ~ . - . . . - . . , < . . . . _ , - . - - . . _ .

SYSTEM 80+= TABLE 2.5.2-1 ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the 1. Inspection of the as-built ESF-CCS 1. For the components and equipment ESF-CCS is as shown on Figure configuration will be conducted. shown on Figure 2.5.2-1, the as-built 2.5.2-1. ESF-CCS conforms with the Basic Configuration.

2. Each division of the ESF-CCS has the 2. Inspection of the four as-built ESF-CCS 2. Each ESF-CCS division has equipment following elements, as depicted on divisions will be performed. for the following:

Figure 2.5.2-2:

selective 2-out-of-4 logic, selective 2-out-of-4 logic, component control logic, component contml logic, process instmmentation, process instrumentation, signal conditioning equipment, signal conditioning equipment, maintenance and test panel, maintenance and test panel, control and display interface devices, control and display interface devices, and a master transfer switch. and a master transfer switch.

3. 'the four ESF-CCS divisions are 3. Inspection for separation and isolation of 3. Physical separation exists between the 4 physically separated and electrically the four as-built ESF-CCS divisionswill ESF-CCS divisions. Electricalisolation isolated. be conducted. devices are provided at interfaces between the four ESF-CCS divisions.

4. Each ESF-CCS division is powered 4. Tests will be performed on the ESF- 4. Within the ESF-CCS, a test signal exists from its respective Class IE bus. CCS by providing a test signal in only only at the equipment powered from the one Class IE bus at a time. Class IE bus under test.

2.5.2 06-18-93 l

l

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

5. Each ESF-CCS division receives 4 chan- 5. Tests will be performed using simulated 5.a) Each ESF-CCS division receives four nels of initiation signals from the PPS PPS signals for ESF initiation input to channels of PPS initiation signals for which are processed using selective 2- each division of the ESF-CCS. each ESF actuation function performed out-of-4 logic to generate actuation sig- by that ESF-CCS division.

nals for the ESF systems controlled by that division. Basic block diagrams for the function logic used in the ESF-CCS for actuation of ESF systems are shown on Figures 2.5.2-3 and 2.5.2-4.

5.b) For each ESF actuation function per-formed by an ESF-CCS division, receipt of an ESF initiation signal from only one PPS channel does not result in gen-eration of an ESF actuation signal and receipt oflike initiation signals from two or more PPS channels does result in generation of an ESF actuation signal for that ESF function.

2.5.2 06-18-93

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Desien Commitment InSDections. Tests. Analyses AcceDiance Criteria

6. He ESF-CCS provides control cap- 6.a) Tests will be performed on the 6.a) He control and display interface ability and, upon receipt of initiation as-built ESF-CCS control and display equipment provide control capability for signals from the PPS, automatically gen- interface equipment. the following systems:

erates actuation signals to the following ESF systems within allocated response safety injection system, times: containment isolation system, containment spray system safety injection system, main steam isolation, and containment isolation system, emergency fealwater system.

containment spray system main steam isolation, and emergency feedwater system.

6.b) Tests will be performed using signals 6.b) ESF initiation signals which satisfy the simulating ESF initiation to the ESF- selective 2 out of 4 criteria result in CCS. actuation signals for related system components for the following systems:

safety injection system, containment isolation system, containment spray system main steam isolation, and emergency feedwater system.

6.c) Tests will be performed using signals 6.c) Measured response times are less than simulating ESF initiation to the ESF- or equal to the response time values CCS. required for ESF actuations.

l l

l 2.5.2 - 06-18-93

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

7. The ESF-CCS provides control cap- 7.a) Tests will be performed on the as-built 7.a) The control and display interface ability and, upon receipt of initiation ESF-CCS control and display interface equipment provide control capability for signals from the PPS, automatically equipment. the following systems:

generates actuation signals to the following non-ESF systems: annulus ventilation system, component cooling water system, annulus ventilation system, onsite power system, component cooling water system, diesel generators, and onsite power system, control complex ventilation system.

diesel generators, and control complex ventilatien system.

7.b) Tests will be performed using signals 7.b) ESF initiation signals which satisfy the simulating ESF initiation to the ESF- selective 2 out of 4 criteria results in CCS. actuation signals for related system components for the following systems:

annulus ventilation system, component cooling water system, onsite power system, diesel generators, and control complex ventilation system.

l 2.5.2 06-18-93

SYSTEM 80+= TABLE 2.5.2-1 (Continucd)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

8. The ESF-CCS provides control cap- 8. Tests will be performed on the as-built 8. The control and ' display interface ability for the following safety-related ESF-CCS control and display interface equipment provide control capability for '

systems: equipment. the following systems:

shutdown cooling system, shutdown cooling system, -

safety depressurization system, safety depressurization system.

atmospheric dump system, atmospheric dump system, station service water system, station service water system, heating, ventilating and air conditioning heating, ventilating and air conditioning systems, and systems, and hydrogen mitigation devices. hydrogen mitigation devices.

9. Upon receipt of ESF initiation signals 9. Tests will be performed using signals 9. Upon receipt of signals simulating for safety injection, containment spray, simulating ESF initiation signals. initiation of safety injection. containment or emergency feedwater, the ESF-CCS spray, or emergency feedwater which initiates an automatic start of the diesel satisfy the selective 2-out-of-4 criteria, generators and automatic load the ESF-CCS willinitiate an automatic sequencing of ESF loads. start of the diesel generators and automatic load sequencing of ESF loads.

3 s

I 1

i ,

2.5.2 06-18-93 t

SYSEM 80+" TABLE 2.5.2-1 (Continucd)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 10.a) Upon detecting loss of power to Class 10.a) Tests will be performed using simulated 10.a) Upon loss of power at a Cass IE bus, .,

IE division buses, the ESF-CCS loss of power to the Class IE buses. signals are generated automatically by automatically initiates startup of the each of two ESF-CCS divisions which respective diesel generrtors, shedding of will:

electrical load, transfer of Class 1E bus connections to the diesel generators, and 1) initiate an automatic start of the sequencing to the reloading of safety- emergency diesel generator related loads to the Class !E bus. associated with that division, 1

2) cause each medium voltage switchgear cistuit breaker to open,

3) cause transfer of the Class IE bus connections to the diesel generator, and

4) sequentially reclose each medium voltage switchgear circuit breaker after the diesel generator has started.

2.5.2 06-18-93

SYSTEM 80+= TABLE 2.5.2-1 (Continued)

ENGINEERED S/SETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment inspections. Tests. Analyses Acceptance Criteria 10.b) Upon ESF actuation, the normal load 10.b) A test will be performed using a 10.b) Upon receipt of the ESF initiation sequence is interrupted and priority is simulated loss of power to the Class 1E signal, the ESF-CCS automatically given to loading the actuated ESF buses and simulated ESF initiation interrupts the loading sequence to load systems and associated safety-related signals input to the ESF-CCS during the the equipment associated with the ESF systems. reloading sequence for each of the equipment associated with the ESF following ESF initiation signals: initiation signal and then resumes the reloading sequence.

safety injection actuation signal, containment spray actuation signal, emergency feedwater actuation signal to steam generator 1, and emergency feeslwater actuation signal to steam generator 2.

I1.a) The ESF-CCS provides an interlock 11.a) Tests will be performed using signals 11.a) Manual control signals input to the ESF-which prevents the ESF-CCS from simulating RCS pressure input to 'he CCS to open the shutdown cooling generating a signal to open the shutdown ESF-CCS. system isolation valves do not result in cooling system isolation valves when the generation of signals to open the valves RCS pressure is above the entry when the ESF-CCS receives signals pressure of the shutdown cooling simulating RCS pressure that is greater system. than the shutdown cooling system entry pressure.

2.5.2 06-18-93

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

- ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Desima Counmitinent Inspections. Tests. Analyses Acceptance Criteria 11.b) The ESF-CCS provides an interlock 11.b) Tests will be performed using signals 11.b) Manual control signals input to the ESFi ,

which prevents the ESF-CCS from gen- simulating RCS pressure input signals to CCS to close the SIT isolation valves do '!

erating signals to close the SIT isolation the ESF-CCS. not result in generation of signals to valves when the RCS pressure is above close the valves when the ESF-CCS re-the entry pressure of the SCS. ceives signals simulating RCS pressure that is greater than - the ' SCS ' entry pressure.

l 11.c) The interlock on the EFW isolation i1.c) Tests will be performed using signals 11.c) Input of signals indicating high SG level valves automatically closes the isolation simulating SG level and Emergency results in generation of a signal to close valves on high SG levels when an Emer- Feedwater Actuation input signals to the the EFW isolation valves unless signals

, gency Feedwater Actuation Signalis not . ESF-CCS. for Emergency Feedwater Actuation are present. also input to the ESF-CCS.

12. The operator interface devices of the 12. Addressed in 6.s), 7.a) and 8.a). 12. Addressed in 6.a),7.a) and 8.a).

ESF-CCS in the MCR provide for auto-matic and manual control of ESF sys-tems and components.

i

~

i a

4 2.5.2 -8' 06-18-93 t

.m ,-. . . - . -.. _ . _ , . . . - . . . . _ - . . _ . . - . . . _ ,_.w.- w..

,_...,v~,.-,,

SYSTEM 80+" TAllLE 2.5.2-1 (Continued)

ENGINEERED SAFE'IY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Desien Commitment Insocctions. Tests. Analyses Acceptance Criteria

13. In the remote shutdown room, operator 13. Tests will be performed on the 13. Control capability is provided at the interface devices provide for manual as-built ESF-CCS control and display ESF-CCS control and display interface control of ESF system components interface devices in the remote shutdown devices in the remote shutdown room needed to achieve hot standby. room following a transfer of control for the following systems:

capability to the remote shutdown room.

safety injection system, emergency feedwater system, component cooling water system, onsite power system, diesel generators, shutdown cooling system, safety depressurization system, atmosphereic dump system, station service water system, and heating, ventilating and air conditioning systems.

i 14.a) Actuation of master transfer switches at 14.a) Tests will be performed using master 14.a) Upon actuation of the master transfer either exit of the MCR transfers control transfer switches at each exit of the switches at either MCR exit:

capability from the ESF-CCS control MCR and each of the ESF-CCS control and display interface devices depicted in and display interface devices in the 1) control actions at the ESF-CCS con-the MCR to those in the remote MCR and the rennte shutdown room. trol and display interface devices do shutdown room, not cause the ESF-CCS to generate the associated control signals, and

2) control actions at the ESF-CCS con-trol and display interface devices in the remote shutdown room cause the ESF-CCS to generate the associated control signals.

2.5JE 06-18-93 l

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

_ ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment InsDections. Tests. Analyses Acceptance Criteria 14.b) Each ESF-CCS division's maintenance 14.b) Testing will be performed using each 14.b) Upon actuation of the master transfer and test panel pmvides capability to ESF-CCS division's maintenance and switching function fmm each ESF-CCS transfer control from the MCR to the test panel and control and display division's maintenance and test panel:

remote shutdown panel for its respective interface devices in the MCR and the ESF CCS division and to transfer con- remote shutdown room. 1) control actions at the ESF-CCS con-trol back to the MCR for its respective trol and display interface devices in ESF-CCS division. the MCR for that ESF-CCS division do not cause the ESF-CCS to gener-ate the associated control signals, and

2) control actions at the ESF-CCS con-trol and display interface devices in the remote shutdown room for that ESF-CCS division cause the ESF-CCS to generate the associated con-trol signals.

2.5.2 06-18-93 i

l l

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

ENGINEERED SAFEIY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria l

Design Commitment Inspections. Tests. Analyses .A_cceptance Criteria l

14.b) (Continued)

Upon de-actuation of the master transfer switching function from each ESF-CCS

( division's maintenance and test panel: t

3) control actions at the ESF-CCS con-trol and display interface devices in the remote shutdown room for that ESF-CCS division do not cause the ESF-CCS to generate the associated control signals, and

4) control actions at the ESF-CCS con-trol and display interface devices in the MCR for that ESF-CCS division cause the ESF-CCS to generate the associated control signals.

l l

j 2.5.2 06-18-93

- , . , . . . . . . . . . - . - - . . . . . ~ . . . - . _ - . - - - . . - - - . - . - . - -

SYSTEM 80+" TAllLE 2.5.2-1 (Continued)

ENGINEERED SAFirlY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 15.a) Diverse manual actuation switches are 15.a) Tests will be performed using the 15.a) Actuation of the switches provides provided as an attemate means for diverse manual actuation switches. signals to achieve actuation of ESF manual actuation of ESF components in components for the following:

two divisions of the ESF-CCS as follows:

2 trains of safety injection, 2 trains of safety injection, I train of containment spray, I train of containment spray, I train of emergency feedwater to each I train of emergency feedwater to each steam generator steam generator 1 main steam isolation valve in each I main steam isolation valve in each main steam line, main steam line, 1 isolation valve in each containment air 1 isolation valve in each containment air purge line, and purge line, and I letdown isolation valve.

I letdown isolation valve.

15.b) The diverse manual actuation switches 15.b) Inspection will be performed on the as- 15.b) Communication of the signals from the provide signals to the lowest level in the built ESF-CCS equipment and, if digital diverse manual actuation switches ESF-CCS digitalequipment. Communi- equipment is used to communicate these implements either:

cation of the signals from the switches is signals, inspection of the design docu-diverse from the software used in the mentation will be performed to confirm 1) hardwired signal communication; or higher levels of the ESF-CCS. that the softwam was developed by a different design group than used to 2) digital equipment that uses micro-develop the ESF-CCS software. processors which are diverse from the microprocessors used in the PPS and ESF-CCS equipment and soft-ware for which the design documen-tation confirms that the software was developed by a different design group than the group (s) which developed the PPS and ESF-CCS software.

2.5.2 06-18-93

SYSTEM 80+" TABLE 2.5.2-1 (Continucd)

ENGINEERED SAFFIIY FEATURES COMPONENT CONTROL SYSTEM Inspections. TcSts. Analyses and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 15.c) Actuation of the switches provides a 15.c) Tests will be performed for each diverse 15.c) Each diverse manual actuation switch is signal which overrides the higher level manual actuation switch with concurrent able to generate a signal which overrides signals, to actuate the associated ESF and opposing control commands initiated the manual signals input via the control component or components. from the control and display interface and display interface devices, such that devices depicted on Figure 2.5.2-2. signals are provided to the associated motor control centers to actuate the ESF equipment.

16.a) Periodic testing to verify operability of 16.a) Inspection of design documentation will 16.a) ne design documentation specifies tests the ESF-CCS can be performed with the be performed to verify the capability to that can be performed while the plant is reactor at power or when shutdown perform eurveillance tests while the operating without disabling the without interfering with the protective plant is operating. protection functions to verify operability function of the system. of the selective 2-out-of-4 logic and the resTonse of ESF systems to ESF actuation signals and interlocks.

16.b) Capability is provided for testing all 16.b) Inspection of the as-built ESF-CCS 16.b) Testing capability provides overlap in functions from ESF initiating signals equipment will be performed to verify individual tests such that all functiens received from the PPS through to the the capability for functional testing. from ESF initiating signals received actuation of protective system from the PPS through to the actuation of equipment. protective system equipment are tested.

16.c) The maintenance and test panel provides 16.c) Inspection of the as-built ESF-CCS 16.c) The maintenance and test panel includes capability for manual testing of ESF- equipment will be performed. the capability to perform manual testing CCS functions and hardware. of ESF-CCS functions and hardware.

i 2.5.2 06-18-93

SYSTEM 80+" TAHLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM InsDections. Tests. Analyses and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

17. Where the ESF-CCS and the process 17. Inspection of the as-built ESF-CCS 17. Electrical isolation devices are provided control system interface to the same configuration will be conducted. between the process control system and component, electrical isolation devices sensors, signal conditioners and actuated are provided between the process devices which interface to the ESF-control system and the shared CCS.

component.

18. Electrical isolation devices are provided 18. Inspection of the as-built ESF-CCS 18. Electrical isolation devices are provided at ESF-CCS interfaces with the DIAS- equipment will be conducted, at ESF-CCS interfaces with the DIAS-N, the DPS, the P-CCS, the control and N, the DPS, the P-CCS, the control and display interface devices, the master display interface devices, the master transfer switcher, and between the signal transfer switcher, and between the signal conditioning equipment and one of the conditioning equipment and one of the two channels of the DIAS-P, as shown two channels of the DIAS-P, as shown on Figure 2.5.2-2. on Figure 2.5.2-2.

19.a) Setpoints for interlocks and actuation or 19.a) Inspection will be performal of the 19.a) The setpoint calculations confirm the use ESF-CCS safety-related functions are setpoint calculations. of setpoint methodologies that require:

determined, documented, installed and maintained according to setpoint

• Documentation of data, assumptions, methodologies which specify and methods used in the bases for requirements for documenting the bases selection of trip setpeints; for selection of trip setpoints, accounting for instrument uncertainties and drift, testing of instrumentation setpoint response and replacement of setpoint-related instrumentation.

2.5.2 06-18-93

___ __-______ _--_ _____.____ - -_______ -__. _ -- ~. _ . . _ _ _ _ ._ ._ _ _ _ _ _ _ . _ _ _ _ . . . . . -

SYSTEM 80+= TABLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES CO'k1LT]NENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 19.a) (Continued)

• Consideration of instrument calibration uncertainties and uncertainties due to environmental conditions, instrument drift, power supply variation and the effect of design basis event transienus is included in determining the margin between the trip setpoint and the safety limit;

• The methods used for combirdng uncertainties;

• Use of written procedures for preoperational testing and tests performed to satisfy Technical Specifications; and

• Documented evaluation for equivalent or better performance of replacement instmmentation which is not identical to the original equipment.

I l

19.b) The setpoint calculations are consistent 19.b) Inspection will be performed of the as- 19.b) The configuration of the as-built in-with the physical configuration of the built instrumentation. strumentation is consistent with the instmmentation. attributes used in the setpoint cal-I culations for location of taps and sensing lines, slope, seismic restraints and pro-i visions for venting, draining, equalizing and process isolation.

2,5.2 06-18-93

SYSTEM 80+" TAHLE 2.5.2-1 (Continucd)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM InsDections. Tests. Analyses and Acceptance Criteria Design Commitmen_t Inspections. Tests. Analyses Acceptance Criteria

20. ESF-CCS software is dedgned, tested, 20. Inspection will be performed of the 20.a) The process defines the organization, installed and maintained using a process process used to design, test, install, and responsibiinies and activities for the maintain the ESF-CCS software. following phases of the software engi-a) which defines the organization, res- neering life cycle:

ponsibilities, and activities for the software engineering life cycle and

• Establishment of plans and method-ologies for all software to be developed; b) which specifies requirements for software quality assurance, veri-

• Specification of functiord, system and fication and validation, configuration software requirements and identification management, and operations and of safety critical requirements; maintenance, and

• Design of the software architecture, pro-c) which incorporates a graded ap- gram structure and definition of the soft-proach according to the software's ware modules; relative importance to safety.

• Development of the software code and testing of the software modules;

• Interpretation of software and hardware and performance of unit and system tests;

• Software installation and checkout testing; and

• Reporting and correction of software defects during operation.

2,5.2 06-18-93

SYSTEM 80+" TABLE 2.5.2-1 (Continued)

ENGINEERED SAFETY FEATURES COMPONENT CONTROL SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 20.b) The process has requirements for the following software development functions:

• Software management, which defines or-ganization responsibilities, documen-tation requirements, standards for soft-

, ware coding and testing, review require-ments, and procedures for problem re-porting and corrective actions;

• Software configuration management, which establishes methods for main-taining historical records of softw3re as it is developed, controlling softwsre changes and for recording and reporting software changes; and

• Verification and validation, which specifies the requirements for the verification review process, review and test activity documentation and reviewer independence.

20.c) The process establishes the method for classifying ESF-CCS software elements according to their relative importance to safety. The process defimes the tasks to be performed for software assigned to each safety classification. ,

2.5.2 06-18-93

e t

SYSTEM 80+"

2.5.4 POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Design Description The Power Control System and the Process-Component Control System (PCS/P-CCS) are non-safety-related instrumentation and control systems which provide control of functions to maintain the plant within its normal operating range for all normal modes of plant operation.

The PCS/P-CCS are located in the nuclear island structures.

The Basic Configuration of the PCS/P-CCS is as shown on Figure 2.5.4-1.

The PCS/P-CCS use sensors, transmitters, signal conditioning equipment, control and display interface devices and digital processing equipment which perform the calculations, data communications, and logic to support the control functions. He digital equipment and software used in the PCS/P-CCS are diverse from that used in the plant protection system (PPS) and the engineered safety features - component control system (ESF-CCS).

We PCS/P-CCS provide control interfaces for the following control functions: 1 reactivity control using control element assemblies, I reactor power cutback, power change limiter, pressurizer pressure and level, main feedwater flow, main steam bypass flow, boren concentration, alternate reactor trip actuation, and alternate emergency feedwater actuation.

nc dreuits used for alternate actuation of reactor trip, turbine trip, and emergency fecowater are indeper, dent and diverse from the protection system actuation circuits.

The PCS/P-CCS provide the following information to the Discrete Indication and Alarm System (DIAS):

1 alt 'rnate reactor trip status, alternate feedwater actuation signal status,  ;

pressurizer pressure, and i steam generator 1 and 2 levels.

2.5.4 -

t l

SYSTEM 80+"

i For parameters used in PCS/P-CCS control functions which are provided from the redundant Class 1E sensors that are used independently by each channel of the protective system, the PCS/P-CCS monitor the four redundant instrument channels.  ;

'Ihe PCS/P-CCS apply signal validation logic to the signals received from the four 1 redundant channels to detect bypassed or failed sensors and to determine the sensed value to be used in the control system. l l

Control and display interface devices for the PCS/P-CCS are provided in the main l control room (MCR) and in the remote shutdown room for control and monitonng i of PCS/P-CCS controlled equipment.  !

Actuation of master transfer switches at either exit of the MCR transfers control capability from the PCS/P-CCS control and display interface devices in the MCR to those in the remote shutdown room. The transfer can also be performed at the  !

PCS/P-CCS equipment cabinets, which also provide capability for transferring control back to the MCR. i Electrical isolation devices are implemented between the PCS/P-CCS and the protection system signal conditioning equipment for each protection signal provided  :

to them, as shown on Figure 2.5.4-2. Electrical isolation devices are provided for the i PCS/P-CCS interfaces with the MCR equipment, the remote shutdown room  ;

equipment, the DIAS and the Data Processing System (DPS), the protection system, i and with protection system components as shown on Figure 2.5.4-2. l

. The PCS/P.CCS software is designed, tested, installed and maintained using a process j which defines the organization, responsibilities, and activities for the software quality i assurance, verification and validation, configuration management, and operations and {

maintenance, and which incorporates a graded approach according to the software's  ;

relative importance to safety. l Inspection, Test, Analyses, and Acceptance Criteria l Table 2.5.41 specifies the inspections, tests, analyses and associated acceptance  !

criteria for the Power Control System / Process-Component Control System.

l l

i 1

i i

i  !

2.5.4 l l

~

i l

SYSTEM 80 +

l 1 I I PCS/P-CCS I I CONTROL & DISPLAY I I INTERFACE DEVICES I I i 1 g MASTER TRANSFER I SWITCHES COMPONENT I CONTROL i

~~~~~~~~~~~~~~~~l LOGIC g i l_. I g y DIAS & DPS g  ;

_. .' I g y ,

I l-PROTECBON I [

l SYSTEM I y SIGNAL VALIDATION I l l I I I l I SIGNAL I-I I i

_ CON. DI.TI.ONING y ,

I I i_____..___ .

I

___1___

em i PROTECTION I (Sj y SYSTEM l ,% --A ---

. COMPONENTS , I

_ _ _ (sjI PROCESS

-- ICOMPONENTS I FIGURE 2.5.4-1 PCS/P-CCS CONFIGURATION

SYSTEM 80 +

l MAIN CONTRO5 ROOM I KEY:

I CONTROL & DISPLAY 4 ==--O-88----- >  ? HARDWIRED OR g

INTERFACE DEVICES DATA UNK i omR TRANSFw SW CH

........e....... . ---Mgcaog=

g DISCRETE SIGNAL (E.G. FIBER OPTIC) lpeupre SauToOG ROOu- 'l PCS/P-CCS POWER SuePt.Y I C COMPONENT S"T,"n'oT&T! 4 ----D-------> CONTROL El ISOLA M l - - - - -

LOGIC DIAS T>GCM P:MATION &

! DATA PROCESSING SYSTEM I4 ...,,g .....,,,,,,, J LARM SYSTEM b ----- 8 MCC MOTOR CONTROL

=--==-

CEhTERS l= DIAS - CHANNEL N """" h """""""""

w ,,,, _ ,,,,,,,,,,I RTSG REACTOR UIP g- ps- _

SWITCHGEAR

---

a ji JL e ji JL PROTECTION SYSTEMS ' CEDMCS CONTROL s ELEMENT DRNE l - - " "

!= se m@ = = = = m uf g g MECHANISM SIGNAL g SIGNAL CONTROL SYSTEM l l"" SIGNAL P==@""--" % VAUDATION y a CONDITIONING E

qy gCONDITIONING

b. --

jL I CEDMCS POWER j l~ ~' ~ l l~ WIRED $ HARDWIREDOR SUPPLY MCC e"OFc, GATE

,..g -g l l 1r r 5 In1SG I- '

\s/ i j.

e 8

= ,- ,

l- I S)

--.s. --

E o e as e e e as sh a as a sa w se as e as as um as as um su a as as um as un su as e e as mE lIII'-

l linREDI l .

....a "

u i MCC" l l

l 1 r

, l l~TURBINEl f -y l EMERGENCY FEEDWATER l

. ,, .C,,ONTJOL_S,,, $M.P,S,, & VAL 3S,,,, ,

FIGURE 2.5.4 i PCS/P-CCS AND INTERCONNECTIONS

- _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ _ _ _ _ _ _ _ _ ~

SYSTEM 30+= TABLE 2.5.4-1 POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Insocctions. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the PCS/P- 1. Ingection of the as-built configuration 1. For the components and equipr.mt CCS is as shown on Figure 2.5.4-1. of the PCS/P-CCS will be conducted, shown on Figure 2.5.4-1, the as-built PCS/P-CCS conforms with the Basic Configuration.

?. The digital equipment and software used 2.a) Inspection of the as-built PCS/P-CCS, 2.a) He digital equipment used in the in the PCS/P-CCS are diverse from that PPS and ESF-CCS equipment will be PCS/P-CCS uses microprocesson which used in the PPS and ESF-CCS. performed. are diverse from the microprocessors used in the PPS and ESF-CCS equipment.

2.b) Inspection of the design documentation 2.b) He software documentation confirms will be performed to confirm that the that the design group (s) which developed software was developed by different the PCS/P-CCS software is different design groups. from the design group (s) which developed the PPS and ESF-CCS software.

3. The PCS/P-CCS pmvide control inter- 3 Inspection will be performed on the as- 3. PCS/P-CCS control interfaces are pro-faces for the following control functions: built PCS/P-CCS control interface vided for the following functions:

equipment, reactivity control using control element reactivity control using control element assemblies, assemblies, reactor power cutback, reactor power cutback, power change limiter, power change limiter, pressurizer pmssure and level, pressurizer pressure and level, main feedwater flow, main feedwater flow, steam bypass flow, steam bypass flow, boron concentration, boron concentration, alternate reactor trip actuation, and alternate reactor trip actuation, and alternate emergency feedwater actuation. alternate emergency feedwater actuation.

2.5.4 _ _ ______ _ _ _ _ --- _.. __ - .-

SYSTEM 80+= TABLE 2.5.4-1 (Continued)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, tests, AnalyScS, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

4. He circuits used for alternate actuation 4. Inspection of the design documentation 4. The documentation confirt::s that circuits of reactor trip, turbine trip and emer- will be performed to confirm that the are implemented in the PCS/P-CCS to gency feedwater are independent and specified alternate actuation circuits are perform actuation of reactor trip, turbine diverse from the protection system independent and diverse from the trip and emergency fee:! water which do actuation circuits, protection system actuation circuits. not utilize signals frvm the PPS or ESF-CCS and that tne PPS and ESF-CCS digital equipment is not used to communicate the actuation signals from the PCS/P-CCS to the actuated components.

5. He PCS/P-CCS provide the following 5. Inspection will be performed of the as- 5. He following information is available at information to the DIAS: built DIAS equipment. a DIAS-N display device:

alternate reactor trip status, altemate reactor trip status, alternate feedwater actuation signal alternate feedwater actuation signal status, status, pressurizer pressure, and pressurizer pressure, and steam generator 1 and 2 levels. steam generator 1 and 2 levels.

6. For parameters used in PCS/P-CCS con- 6. A test will be performed using signals 6. For each parameter, the representative trol functions which are provided from simulating each parameter provided to parameter value determined by the the redundant Class 1E sensors that are the PCS/P-CCS via the redundant Class PCS/P-CCS from the Class IE sensor used independently by each channel of IE sensors that are used independently inputs is bounded by the three signals the protective system, the PCS/P-CCS by each channel of the protective which are simulated to be unaffected by monitor the four redundant instrument system. The signals will simulate a the failure.

channels. He PCS/P-CCS apply signal failure of one of the four sensor inputs validation logic to the signals received for each parameter.

from the four redundant channels to de-tect bypassed or failed sensors and to determine the sensed value to be used in the control system.

i 2.5.4 . . . ~ - .-_ _ _. , _ ._ __ _

s SYSTEM 80+= TABLE 2.5.4-1 (Continued)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections. Tests. Analyses Acceptance Criteria

7. Control and display interface devices for 7. Inspection will be performed of the as- 7. Control and display interface devices for the PCS/P-CCS are provided in the built PCS/P-CCS control and display the PCS/P-CCS are provided in the MCR and in the remote shutdown room. interface devices in the MCR and MCR and in the remote shutdown room.

remote shutdown room.

8.a) Actuation of master transfer switches at 8.a) Tests will be performed using the 8.a) Upon actuation of the master transfer either exit of the MCR transfers control master transfer switches at each exit of switches at either MCR exit:

capability from the PCS/P-CCS control the MCR and each of the PCS/P-CCS and display interface devices in the control and display interface devices in 1) control actions at the PCs/P-CCS MCR to those in the remote r.hutdown the MCR and the remote shutdown control and display interface devices room. panel. in the MCR do not cause the process control systems to generate the associated control signals; and

2) control actions at the PCS/P-CCS control and display interface devices in the remote shutdown room cause the process control systems to generate the associated control signals. ,

i t

i t

i i

2.5.4  !

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ . _ . __ .. . _ _ . . _ . . . - . . . _ . , . , . . _ . . - . _ .u . .-.. , - , - .-, _ _ _ . . , _ .

t R

SYSTEM 80+" TABLE 2.5.4-1 (Continued)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, Tests, Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acccotance Criteria 8.b) ne transfer of control capability can 8.b) Testing will be performed at the 8.b) Upon actuation of the master transfer also be performed at the PCS/P-CCS equipment cabinets for the PCS/P-CCS switching function from the equipment equipment cabinets, which also provide and using the PCS/P-CCS control and cabinets for the PCS/P-CCS:

capability for transferring control back display interface devices in the MCR to the MCR. and the remote shutdown room. 1) control actions at the PCS/P-CCS control and display interface devices in the MCR do not cause the process control systems to generate the associ(*ti control signals; and

2) control actions at the PCS/P{CS control and display interface devices ,

in the remote shutdown room cause the process control systems to generate the associated control signals.

Upon de-actuation of the master transfer switching function from the equipment cabinets for the PCS/P-CCS:

1) control actions at the PCSIP-CCS control and display interface devices in the remote shutdown room do not cause the process control systems to generate the associated control signals; and

2) control actions at the PCS/P-CCS control and display interface devices in the MCR cause the process con-trol systems to generate the associated control signals.

2.5.4 - - .

SYSTEM 80+" TABLE 2.5.4-1 (Continued)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, Tests, Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 9.a) Electrical isolation devices are provided 9.a) Inspection of the as-built PCS/P-CCS 9.a) Electrical isolation devices are provided between the PCS/P-CCS and the pro- configuration will be conducted. between the PCS/P-CCS and the pro-tection system signal conditioning equip- tection system signal conditioning equip-ment for each protection signal provided ment, consistent with Figure 2.5.4-2 for to them, as shown on Figure 2.5.4-2. each protection signal provided to them.

9.b) Electrical isolation devices are provided 9.b) Inspection of the as-built PCS/P-CCS 9.b) Electrical isolation devices are provided for the PCS/P-CCS interfaces with the configuration will be conducted. for the PCS/P-CCS interfaces with the MCR, the remote shutdown room, the MCR, the remote shutdown room, the safety related display instrumentation, safety related display instrumentation, the protection systems and with pro- the protection systems and with pro-tection system components, as shown on tection system components, conforming Figure 2.5.4-2. to Figure 2.5.4-2.

10. PCS/P-CCS software is 10. Inspection will be performed of the 10.a) The process defines the organization, process used to design, test, install, and responsibilities and activities for the a) designed, tested, installed and main- maintain the PCS/P-CCS software, following phases of the software tained using a process which defines engineering life cycle:

the organization, responsibilities, and activities for the software

• Establishment of plans and engineering life cycle and methodologies for all software to be developed; b) which specifies requirements for software quality assurance, veri-

• Specification of functional, system and fication and validation, configuration software requirements and identification management, and operations and of safety critical requirements; maintenance, and l

c) which incorporates a graded ap-proach according to the software's - (Continued)

! relative importance to safety.

2.5.4 l

SYSTEM 80+" TABLE 2.5.4-1 (Continued)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, Tests, Analyses, and Acceptance Criteria Desinn Commitment inspections. Tests. Analyses Acceptance Criteria 10.a) (Continued)

• Design of the software architecture, pro-gram structure and definition of the soft-ware modules;

• Development of the software code and 5 testing of the software modules;

• Interpretation of software and hardware and performance of unit and system tests;

• software installation and checkout testing; and

• Reporting and correction of software de-fects during operation.

10.b) The process has requirements for the following software development functions:

• Software management, which defines or-ganization responsibilities, docu-mentation requirements, standards for software coding and testing, review re-quirements, and procedures for problem reporting and corrective actions;

. (Continued) 2.5.4 _____ ______ __ _____________________ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

SYSTEM 80+" TAHLE 2.5A-1 (Continned)

POWER CONTROL SYSTEM / PROCESS-COMPONENT CONTROL SYSTEM Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 10.b) (Continued) e Software configuration management, which establishes methods for main-taining historical records of software as it is developed, controlling software changes and for recording and reporting software changes; and

• Verification and validation, which

' specifies the requirements for the veri-fication review process, the validation testing process, review and test activity documentation and reviewer inde-pendence.

10.c) The process establishes the method for classifying PCS/P-CCS software ele-i ments according to their relative irnpor-tance to safety. The process defines the tasks to be performed for software assigned to each safety classi-fication.

t 2.5.4 _ ___. __ _ ___ _ _ - - _ _ ___ _ _ _ __ . _. _. . - .

i SYSTEM 80+"

2.6.1 ELECTRICAL POWER DISTRIBUTION SYSTEM  ;

I Design Description  :

The Electrical Power Distribution System (EPDS) is the onsite power system which distributes electrical power to plant equipment. The EPDS has safety-related and non-safety-related subsystems. The safety-related EPDS subsystem is Class 1E and the non-safety-related EPDS subsystem is non-Class 1E.

The Basic Configuration of the EPDS is as shown on Figures 2.6.1-1 and 2.6.1-2.  ;

I EPDS Non-Class 1E Subsystem The EPDS Non-Class 1E Subsystem consists of the Main Generator (MG), the Generator Circuit Breaker (GCB), and a Unit Main Transformer (UMT) with an installed, single- phase spare transformer, two 3-phase Unit Auxiliary Transformers (UATs) with one installed spare UAT, two 3-phase Reserve Auxiliary Transformers l (RATS), isolated and non-segregated phase buses, load centers, switchgear, breakers, I and cabhng.

I The GCB is sized to interrupt the maximum calculated fault current occurring at the breaker. The GCB is equipped with redundant trip coils supplied from separate DC  !

power systems.

The isolated phase buses from the MG are rated to carry the MG full load output and sustain the maximum fault current until the fault is cleared. l l

Each UAT provides the power requirements of its assigned electrical loads without exceeding its self-cooled rating.

Each RAT provides the power requirements of its assigned electrical loads on the Class 1E buses, the permanent non-safety bus, and at least one reactor coolant pump i and the pump support loads. Each RAT provides these power requirements without exceeding the RAT's self-cooled rating.

The UMT and the UATs are physically separated from the RATS. The UATs are physically separated from each other and from the UMT.

Outside the main control room (MCR), the 7 IG, GCB, UATs, UMT, and associated instrumentation and control circuits are sepat sted from the RATS' instrumentation l and control circuits by a minimum of 50 feet, or by walls or floors, except at the non- {

Class 1E DC power sources where they are routed in separate raceways. UAT and UMT instrumentation and control circuits within the MCR are separated from the RATS

• instrumentation and control circuits by routing the circuits in separate raceways. At least 50 feet separation is maintained between the UATs' output power 2.6.1 06-18-93 i 1

SYSTEM 80+"

l feeders and instrumentation and control circuits and the RATS' output power feeders I and instrumentation and control circuits, except at the medium voltage switchgear l where circuits are routed to the opposite ends of the switchgear. ]

Each UAT and its associated instrumentation and control circuits are separated from each other UAT and its associated instrumentation and control circuits by a minimum of 50 feet, or by walls or floors. Each UAT receives non-Class 1E DC power from separate non-Class 1E batteries.

EPDS Non-Class 1E and Class 1E medium voltage switchgear and low voltage load i center switchgear, with their respective transformers, and the low voltage motor control centers (MCCs) are sized to supply their load requirements. Medium voltage and low voltage load center switchgear, and the MCCs are rated to withstand the maximum fault currents for the time required to clear the fault from the power source.

Medium voltage switching devices may be closed by manual action without their control power.

Non-Class 1E and Class 1E electrical circuits which are routed through containment penetration assemblies are provided with redundant protective devices, unless the maximum current available in the circuit is less than the continuous rating of the penetration assembly.

An electrical grounding system is provided for instrumentation, controls, and computer systems, electrical equipment (e.g., switchgear, motors, distribution panels),

and large components / structures (e.g., metal structures, metallic tanks).

Lightning protection is provided for specified plant structures, transformers, and equipment located outside buildings.

Each grounding system and lightning protection system is separately grounded to the plant ground grid.

EPDS Class 1E Subsystem The EPDS Class 1E Power Subsystem distributes power to safety.related equipment I and associated Class 1E circuits. Non-Class 1E loads which are connected to a Class IE power source are treated as associated Class 1E circuits. The EPDS Class 1E Subsystem has three Class 1E functional subsystems: an AC safety power system, an AC vital instrumentation and control power system, and a DC vital instrumentation and control power subsystem.

2.6.1 06-18-93 1

~

i l

l l

SYSTEM 80+" l 1

For the EPDS, Class 1E power is supplied by two independent Gass 1E Disisions. I Electricalindependence is maintained between Cass 1E Divisions, and between Cass )

1E Divisions and non-Class 1E equipment.

Class IE cables and raceways are identified by a method which specifies DMsional and Channel designations.

Harmonic distortions in non-Cass 1E and Class 1E circuits do not prevent Class 1E equipment from performing its safety functions.

EPDS Class 1E and non-Class 1E interrupting devices are coordinated so that the ,

circuit interrupter nearest to the fault opens before other interrupting devices further j I

from fault.

Class 1E electrical Divisions are physically separated from each other. Class 1E electrical Divisions are physically separated from non-Cass 1E equipment.

Redundant Cass 1E containment electrical penetrations are physically separated and electrically isolated.

Interface Requirements The Offsite Power System has interfaces with the EPDS at Preferred Switchgear Interfaces I and II, as shown on Figure 2.6.1-1. Each Preferred Switchgear Interface consists of an electrically independent and physully separate source of offsite power.

'Ihe Offsite Power System and the interface between the Offsite Power System and ,

the EPDS satisfy the following requirements:  !

I Transmission system voltage variation is no more than plus or minus 10 )

percent of nominal voltage during steady-state operation; Transmission system frequency variation is within plus or minus 2 Hertz of the nominal grid frequency during recoverable periods of system instability; and Each Preferred Switchyard Interface has relaying, control, and monitoring equipment powered from dedicated low voltage AC and DC power systems which are independent and redundant.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.6.1-1 specifies the inspections, tests, analysis, and associated acceptance criteria for the Electrical Power Distribution System.

2.6.1 06-18-93

PREFERRED SWITCHVARD SYSTEM 80+ TM INTERFACE I PREFERRED -

RFACEI kN SPARE FACEI UPPER MEDfUM

~~

UPPER MEDtUM UPPER MEDIUM

^ ~

N UPPER MEDIUM VOLTAGE VOLTAGE VOLTAGE 40N-SAFETYf) I) d i) *)

] NON-SAFETY {) < -) 5) i} ')NON-SAFETY

{ } NON-SAFETY ] { VOLTAGE X-1 _ X-2 Y-2 Y-1

) ) )

f) his'"

VOLTAGE his:

VOLTAGE NON-SAFETY NON-SAFETY X Y j)p l) i) 1)

pp 3

PERMANENT PERMANENT

{)

, NON-SAFETY j)

CT l.) NON-SAFETY {)

X Y NON-CLASS 1E NON CLASS 1E

_ p _ _ q) _ _ _ _ _ )_ , )_. _ _ _ _ _ _ _ _ _ y, , _ _ . .) _ _ _ _ q) _ _ _ pg,,3 SAFETY $

SAFETY

5) 1) SAFETY })

BUS D

{) BUS B BUS A } } BUS C SAFETY

!,) b FIGURE 2.6.1-1 AC ELECTRICAL POWER DISTRIBUTION SYSTEMS

SYSTEM 80 +"

SAFETY BUS C SAFETY BUS A SAFETY BUS A CLASS 1E CLASS 1E CLASS 1E 480 V MCC 480 V MCC 480 V MCC CHANNELC - CHANNEL A DIVISION 1I 1r 125V DC BATTERY lI 12SV DC BATTERY 125V DC BATTERY CHANNEL C E CHANNEL A Z DMSION C

- BATTERY -

BATTERY BATTERY CHARGER CHARGER l s CHARGER I s( I s( -

CHANNEL C 125V DC CHANNEL A 125V DC DMSiON 125V DC [

DISTRIBUTION CENTER DISTRIBUTION CENTER DISTRIBUTION CENTER [

( ( l CHANNEL C CHANNEL A DMSION 120V AC AUXILIARY 120V AC AUXIUARY 120V AC AUXILIARY POWER INVERTER POWER INVERTER POWER INVERTER

) ) )

VITALi&C CHANNEL C VITALI&C CHANNEL A VITAll&C DIVISION 120V AC DISTRIBUTION - 120V AC DISTRIBUTION 120V AC DISTRIBUTION CENTER CENTER CENTER FIGURE 2.6.1-2 i CLASS 1E DC AND VITAL AC INSTRUMENTATION

AND CONTROL POWER SUPPLY SYSTEM (ONE OF TWO DIVISIONS)

-, .._ ,.m.,- , . , . , - .

<, . .~_ _ _ _ _ _ _ . _ _ _ .

SYSTEM 80+" TABLE 2.6.1-1 ELECTRICAL NMVER DISTRIBUTION SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acccotance Criteria

1. The Basic Configuration of the EPDS is 1. Inspction of the as-built EPDS 1. For the components and equipment as shown on Figures 2.6.1-1 and configuration will be conducted. shown on Figures 2.6.1-1 and 2.6.1-2, 2.6.1-2. the as-built EPDS conforms with the Basic Configuration.

2.a) The GCB is sized to interrupt the 2.a) An analysis will be performed of the 2.a) The GCB nameplate rating is greater maximum calculated fault current fault current intenupt capability of the than the maximum calculated fault occurring at the breaker. GCB. current occurring at the breaker.

2.b) The GCB is equipped with redundant 2.b) Tests will be performed on the GCB DC 2.b) Within the GCB, a test signal exists trip coils supplied from separate DC power systems by providing a test signal only in the trip circuit under test.

power systems. in only one GCB trip circuit at a time.

3. The isolated phase buses are rated to 3. An analysis will be performed for the 3. The isolated phase bus nameplate ratings carry the MG full load output and ratings of each isolated phase bus. show that the MG fullload output can sustain the maximum fault current until be carried and the maximum fault the fault is cleared. current can be sustained until the fault is cleared.

4. Each UAT provides the power 4. An analysis will be performed for the 4. Load analyses for the as-built UATs requirements of its loads without load requirements for each UAT. exist and conclude that each UAT's exceeding its self-cooled rating. capacity, as determined by its nameplate self-cooled rating, exceeds its analyzed load requirements.

2.6.1 06-18-93

SYSTEM 80+ TABLE 2.6.1-1 (Continued)

ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

5. Each RAT provides the power 5. An analysis will be performed for the 5. l. cad analyses for as-built RATS exist requirements of its assigned electrical load requirements for each RAT. and conclude that each RATS' capacity, loads on the Class IE buses, the as determined by its nameplate self-permanent non-safety bus, and at least cooled rating, exceeds its analyral load one reactor coolant pump and the pump requirements.

support loads. Each RAT provides these power requirements without exceedmg the RAT's self-cooled rating.

6.a) ne UMT and the UATs are physically 6.a) Inspection will be performed of the as- 6.a) He UMT and UATs are separated from separated from the RATS. built UMT, UAT and RAT installations. the RATS by a minimum of 53 feet.

6.b) The UATs are separated from each 6.b) Inspection will be perfonned of the as- 6.b) ne UATs are separated from each other and from the UMT. built UMT and UAT installations. other and separated from the UMT by a minimum of 50 feet.

6.c) Outside the MCR, the MG, GCB, 6.c) Inspection of the as-built MG, GCB, 6.c) Outside the MCR, the MG, GCB, UATs, UMT, and associated instm- UAT, UMT, and RAT installations will UATs, UMT, and associated instrumen-mentation and control circuits are be performed. tation and control circuits are separated separated from the RATS' instru- from the RATS' instrumentation and mentation and control circuits by a control circuits by a minimum of 50 minimum of 50 feet, or_ by walls or feet, or by walls or floors, except at the floors, except at the non-Class IE DC non-Class IE DC power sources where power sources where they are routed in they are routed in separate raceways, separate raceways.

{

l 2.6.1 06-18-93

_ _ _ . _ _- . - - - - --__ _ _ _ , . ._ _- _ -- __ _ ~_ _ _ _ . _ _ _ _ _ _

SYSTEM 80+ TABLE 2.6.1-1 (Continued)

, ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 6.d) UAT and UMT instrumentation and 6.d) Inspection of the UAT. UMT, and RAT 6.d) UAT and UMT instrumentation and control circuits within the MCR are instrumentation and control circuits control circuits within the MCR are separated from the RATS' within the MCR will be performed separated from the RATS

• instrumentation and control circuits by instrumentation and control circuits by routing the circuits in separate routing the circuits in separate raceways. raceways.

6.e) At least 50 feet separation is maintained 6.e) Inspection of the UAT and RAT output 6.e) At least 50 feet separation is maintained between the UATs' output power power feeders and instrumentation and between the UATs' output powee feeders and instrumentation and control control circuits will be performed. feeders and instrumentation and control circuits and the RATS' output power circuits and the RATS

• output power feeder and instrumentation and control feeders and instrumentation and control circuits, except at the medium voltage circuits, clicept at the medium voltage switchgear where circuits are routed to switchgear where circuits are routed to the opposite ends of the switchgear. the opposite ends of the switchgear.

6.f) Each UAT and its associated 6.f) Inspection of the as-built UATs will be 6.g) Each UAT and its associated instrumentation and control circuits are performed. instrumentation and control circuits are separated from each other UAT and its separated from each other UAT and its associated instrumentation and control associated instrumentation and control circuits by a minimum of 50 feet, or by circuits by a minimum cf 50 feet, or by walls or floors, walls or floors.

6.g) Each UAT receives non-Class IE DC 6.g) Inspection of the as-built UAT power 6.g) Each UAT receives non-Class IE DC power from separate non-Class IE supplies will be performed. power from separate non-Class IE batteries. batteries.

2.6.1 06-18 93

SYSTEM 80+ TABLE 2.6.1-1 (Continued _1 ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 7.a) EPDS non-Class 1E and Class 1E 7.a) Analyses for the as-built equipment 7.a) Load analyses for the as-built equipment medium voltage switchgear and low loading will be performed. Tests will exist and conclude that the capacities of voltage load center switchgear, with be performed with connected Class IE the Class IE electrical equipment, as their respective transformers, and the loads operated at 9% to 10% above and determined by their nameplate ratings, low voltage MCCs, are sized to supply 9% to 10% below design voltage. exceed their analyzed load requirements, their load requirements. Connected Class lE loads operate in the ranges of 9% to 10% above and 9% to 10% below design voltage.

7.b) Medium vcitage and low voltage load 7.b) Analyses for the maximum fault currents 7.b) The analyzed fault currents do not center switchgear, and the MCCs are on as-built breakers, switchgear, motor exceed the interrupt capacity of the rated to withstand the maximum fault control centers, isolated and non- EPDS non-Class IE and Class IE currents for the time required to clear segrtgated phase buses, and cables will breakers, switchgear, motor control the fault from the power source. be performed. centers, isolated and non-segregated buses, and cables.

8. Medium voltage switching devices may 8. A test will be performed to manually 8. Medium voltage switching devices are be closed by manual action without their close medium voltage switching devices verified to be capable of being manually control power. with control power supplies to the tested closed without their control power switching devices deenergized. supplies. .

9. Non-Class IE and Class IE circuits 9. Analyses of each circuit penetrating 9. Redundant overturrent devices are which are routed through containment containment for the maximum available provided on penetration circuits, unless penetration assemblies are provided with current will be performed. Inspection the analyses conclude that the maximum redundant protective devices, unless the of overcurrent devices on cables will be current available in the circuit is less maximum current available in the circuit performed. than the continuous current rating of the is less than the continuous rating of the penetration assembly.

penetration assembly.

2.6.1 06-18-93

SYSTEM 80+ TABLE 2.6.1-1 (Continued)

ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 10.a) An electrical grounding system is 10.a) Inspection of the plant grounding system 10.a) An electrical grounding system exists provided for instrumentation, controls, will be performed. for instrumentation, controls, and and computer systems, electrical computer systems, electrical equipment, equipment, and large compon- and large components / structures.

ents/ structures.

10.b) Lightning protection is provided for 10.b) Inspection of the lighting protection 10.b) Lightning protection is provided for specified plant structures, transformers system will be performed. specified plant stnictures, transformers and equipment located outside buildings. and equipment located outside buildings.

10.c) Each grounding system and the lightning 10.c) Inspection of the plant ground grid will 10.c) Each ground system and the lightning prutection system is separately grounded be performed. protection system have separate ground to the plant ground grid. connections to the plant ground grid.

I1. Non-Class 1E loads which are connected i1. Inspection of as-built EPDS confirms 11. Non-Class 1E loads which are connected to a Class IE power source are treated non-Class IE and Class IE to a Class IE power source are treated as Class IE Associated circuits. ' classifications. as Class IE Associated circuits.

12. Electrical independence is maintained 12. Tests will be performed by providing a 12. A test signal exists only in the Division between Class ' IE Divisions, and test signal in only Class IE Division at under test.

between Class IE Divisions and non- a time.

Class 1E equipment.

13. Class IE cables and raceways are 13. Inspection of Class IE cabling and 13. Color coding of Class IE cabling and identified by a method which specifies raceways will be performed. raceways reflects Divisional and Divisional and Channel designations. Charmel designations.

2.6.1 06-18-93 l

SYSTEM 80+ TABLE 2.6.1-1 (Continued)

ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analyses, and AcccDtance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

14. Hannonic distortions in non-Class IE 14. Harmonic distortion analyses on the as- 14. He analyzed harmonic voltage and Class IE circuits do not prevent built EPDS will be performed. distortion does not exceed 5 percent Class IE equipment from performing its voltage distortion on the Class IE safety function. electrical distribution systems.

15. EPDS Class IE and non-Class IE 15. Analyses of breaker coordination on 15. He breaker coordination analyses interrupting devices are coordinated so as-installed intenupting devices will be specify the interrupting device that the circuit interrupter nearest to the performed. coordination required to interrupt the fault opens before other interrupting fault at the location closest to the fault.

devices further from the breaker.

16. De Class IE Electrical Divisions are 16. Inspection of as-built electrical Divisions 16. He Class IE electrical Divisions are physically separated from each other. will be performed. physically separated by a Divisionalwall Class IE Divisions are physically or a fire barrier, except for components separated from non-Class IE equipment. of the system within Containment and within the MCR which are separated by spatial arrangement or barriers.

Components of the system within the MCR are physically separated and electrically isolated from components in the remote shutdown room.

2.6.1 06-18-93

SYSTEM 80+ TABLE 2.6.1-1 (Continued)

ELECTRICAL POWER DISTRIBUTION SYSTEM Inspections. Tests. Analvscs. and Acceptance Criteria Desien Conimitment Inspections. Tests. Analyws Acceptance Criteria

17. Redundant Class IE containment 17. Inspection of redundant Class IE 17. Redundant Class IE containment electrical penetrations are physically and Containment electrical penetrations will elect:ical penetrations are physically electrically separated. be performed to cmfirm physical separated to maintain the independence separation. Tests will be performed on of Class IE circuits and equipment. For redundant Containment electrical each Cnntainment electrical penetration, penetrations by providing a test signal in a test signal exists only in the Division only one of the redundant Containment under test.

electrical penetrations at a time.

2.6.1 06-18-93

_ _ - _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ - - _ - _ - _ - - _ - - - _ . - _ . -- _ _- - -~ . . . ..

SYSTEM 80+" l 2.6.2 ONSITE STANDBY AC POWER SOURCES Design Description j The Onsite Standby AC Power Sources are used to provide onsite emergency electrical power. The onsite standby AC power sources consist of two emergency diesel generators (EDGs), which are safety-related, and a combustion turbine / generator (CT), which is non-safety-related.

One EDG is connectable to the two Class 1E buses of a Cass 1E Division and the other EDO is connectable to the two Cass IE buses of the other Gass 1E Division. l The CT is connectable to either or both of the permanent non-safety buses.

The EDGs are located in physically separate areas of the nuclear island structures. 1 The diesel fuel storage tanks for each of the two EDGs are located in physically separate diesel fuel storage structures. The CT is located outside the nuclear island structures, outside the turbine low trajectory missile path, and is physically separated i

~

from the EDGs and their support systems.

l The EDGs are sized to supply their load demands following a design basis accident I which requires use of emergency power.

Each EDG has a diesel fuel oil system, a diesel lube oil system, a diesel engine cooling system, a starting air system, and a diesel air intake and exhaust system. I I

Each EDG has fuel storage capacity to provide fuel to its EDG for a period of no less than 7 days with the EDG supplying the power requirements for the most limiting design basis accident.

The starting air system receiver tanks of each EDG have a combined air capacity for 5 starts of the EDG without replenishing air to the receiver tanks.

The EDG combustion air intakes are separated from the EDO exhaust ducts.

Electrical independence is provided between Cass IE Divisions and between the Class 1E Divisions and non-Class 1E equipment.

A loss of power to a Class 1E bus initiates an automatic start of the respective EDG and automatic connection to the affected Class 1E buses in the affected Division.

Each EDG receives an automatic start signal in response to a safety injection actuation signal (SIAS), a containment spray actuation signal (CSAS), or an emergency feedwater actuation signal (EFAS). An EDG does not automatically connect to its Divisional Class 1E buses, if the Divisional Class 1E buses are energized.

2.6.2 06-18-93 l

SYSTEM 80+"

Each EDG starts, attains rated voltage and frequency, and is capable of receiving electrical loads in less than or equal to 20 seconds after receiving a signal to start.

The electricalloads to each EDG are automatically sequenced. Each EDG has the l capacity to start and run its largest motor load at the end of the automatic loadmg j sequence, j When operating in a test mode, an EDG resets to its automatic control mode, if a start signal is received. I EDG equipment protection functions are provided to initiate automatic protective actions for the EDGs. Except for the EDG overspeed, generator differential relay, low-low engine oil pressure, and generator voltage-controlled overcurrent trips, EDG l equipment protection functions are by-passed when an automatic start signal is received by an EDG.

Displays of EDG operating status, voltage, amperage, frequency, engine speeds, starting air pressure, and engine cooling water temperature and pressure instrumentation exist in the main control room (MCR) or can be retrieved there.

Controls exist in the MCR to start, load, and stop each EDG.

Alarms exist in the MCR to indicate EDG protection function bypass and inoperability status.

The CT is a non-Class 1E standby power source which provides power to the EPDS permanent non-safety buses. In response to a loss of power signal, the CT starts automatically within 2 minutes of receipt of a s: art signal and then automatically connects to and loads the permanent non-safety buses.

De CF can also be manually connected to the Class 1E buses.

Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.2-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Onsite Standby AC Power Sources.

2.6.2 06-18-93 l

-. . . . __ -.l

SYSTEM 80+" TAHLE 2.6.2-1 t

ONSITE STANDBY AC POWER SOURCES Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment inspections. Tests. Analyses Acceptance Criteria 1.a) One EDG is connectable to the two 1.a) Inspection of the as-built EDG 1.a) One EDG is cormectable to the two Class IE buses of a Class IE Division configurations will be conducted. Class IE buses of a Class IE Division and the other EDG is connectable to the and the other EDG is connectable to the two Class IE buses of the other Class two Class IE buses of the other Class 1E Division. 1E Division.

1.b) The CT is connectable to either or both 1.b) Inspection of the as-built CT will be 1.b) The CT is connectable to either or both of the permanent non-safety buses. conducted. of the permanent non-safety buses.

2. The EDGs are located in physically 2. Inspection of the as-built EDGs will be 2. The two EDGs are located on opposite separate areas of the nuclear island performed. sides of the nuclear island structures and strue:ures. are separated by the Divisional wall.

3. The diesel fuel storage tanks for each of 3. Inspection of the as-built diesel fuel 3. The diesel fuel storage tanks for one the two EDGs are located in physically storage tank areas will be performed. EDG are located in a different structure separate diesel fuel storage structures. from the diesel fuel storage tanks for the other EDG.

4. The CT is located outside the nuclear 4. Inspection of the as-built Cr wi'l be 4 'Ibe CT is physically separated by island structures, outside the turbine low performed. distance from the EDGs and their trajectory missile path, and is physically support systems and is located outside separated fmm the EDGs and their the turbine low trajectory missile path.

support systems.

5. The EDGs are sized to supply their load 5. Analyses will be performed for the EDG 5. An analysis exists and concludes that demands following a design basis loading profile. each EDG has a capacity based on its accident which requires use of nameplate rating that is greater than or emergency power. equal to its load demands following a design basis accident which requires use of emergency power.

2.6.2 06-18-93

---

,_.----,---_.-------,,---_.-.__-_----_.-._----a.-- --- - - , - - _ - - , - - - - - - _ - - - , - - - - _ _ - - _ . - - , - - - - - - - s + --_m -- - - ~ + ' ~ e

SYSTEM 80+ TABLE 2.6.2-1 (Continuedl ONSirE STANDRY AC POWER SOURCFE Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses . Acceptance Criteria ,

t

6. Each EDG has a diesel fuel oil system, 6. Inspection of the as-built EDG support 6. Each EDG is provided with a diesel fuel a lube oil system, a diesel engine systems will be performed. oil system, a tube oil system, a diesel cooling system, a starting air system, engine cooling system, a starting air and an air intake and exhaust system. system, and an air intake and support system.

7. Each EDG has fuel storage capacity to 7. An inspection and analysis will be 7. An analysis exists and concludes that provide fuel to its EDG for a period of performed to determine fuel storage EDG has fuel storage capacity to

, no less than 7 days with the EDG capacities and EDG fuel consumption. operate the EDG for 7 days with the supplying the power requirements for EDG supplying power during the stost the most limiting design basis accident. limiting design basis accident.

8. The starting air system receiver tanks of 8. Tests will be performed with the EDGs 8. Each EDG can be started 5 times each EDG have a combined air capacity and their air start systems. without replenishing air to the receiver for 5 starts of the EDG without tanks.

replenishing air to the receiver tanks.

. 9. The EDG combustion air intakes are 9. Inspection of the as-built EDG air 9. Each EDG air intake and air exhaust are separated from the EDG exhaust ducts. intakes and air exhaust will be separated by distance and orientation.

performed. The air intakes and exhausts of the two EDGs are separated by the location of the EDGs on opposite sides of the nuclear island structures.

2.6.2 - 06-18-93

SYSTEM 89+ TABLE 2.6.2-1 (Continued)

ONSITE STANDBY AC POWER SOURCES InSDections. Tests. Analyses, and AcceDtimcc Criteria Design Commitment Inspections. Tests. Anahses Acceptance Criteria

10. Electrical independence is provided 10. Tests will be performed on each EDG 10. A test signal exists only in the EDG and between Class IE Divisions and between and support systems by providing a test support systems Division under test.

the Class IE Divisions and non-Class signal in only one Class IE Division at IE equipment, a time.

I1. A loss of power to a Class 1E bus 11. Tests for the actuation and connection of 11. In response to a signal that simulates a initiates an automatic start of the each EDG will be performed using a loss-of-power, the affected EDG respective EDG and automatic signal that simulates a loss-of-power. receives a start signal and is connected connection to the affected Class IE to its Divisional Class IE buses.

buses in the affected Division.

12. Each EDG receives an automatic start 12. Tests for the actuation of each EDG will 12. Each EDG receives a start signal in signal in response to a safety injection be performed using signals that simulate response to each of the following actuation signal (SIAS), a containment a SIAS, a CSAS, and a EFAS. simulated signals; a SIAS, a CSAS, and spray actuation signal (CSAS), or an a EFAS, but does not automatically emergency feedwater actuation signal connect to its Divisional Class 1E buses.

(EFAS). An EDG does not automatically connect to its Divisional Class IE buses, if the Divisional Class IE buses are energized.

13.a) Each EDG starts, attains rated voltage 13.a) A test of each EDG will be performed. 13.a) Each EDG attains its rated voltage and frequency, and is capable of within i 10% and rated frequency receiving electrical loads in less than or within i 2% in less than or equal to 20 equal to 20 seconds after receiving a seconds.

signal to start.

2.6.2 06-18-93

SYSTEM 80+ TAHLE 2.6.2-1 (Continued)

ONSITE STANDHY AC POWER SOURCES InsDections. Tests. Analyses, and AcceDiance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria 13.b) The electrical loads to each EDG are 13.b) Tests of the loading sequence on each 13.b Each EDG is automatically loaded in sequenced automatically. EDG will be perfonned. accordance with its load sequences for accident and for non-accident conditions.

13.c) Each EDG has the capacity to start and 13.c) A test of the loading sequence to each 13.c) Each EDG starts and runs its largest run its largest motor load at the end of EDG will be performed. A failure to motor load after other loads are the automatic loading sequence. load the largest motor load in sequence sequenced on the EDG.

will be simulated.

14. When operating in a test mode, an EDG 14. A test will be performed with each EDG 14. When in a test mode configuration, each resets to its autornatic control mode, if in a test mode configuration. An EDG resets to its automatic control a start signal is received. actuation signal will be simulated. mode upon receipt of a simulated start signal.

15. Except for the EDG overspeed, 15. Tests will be performed on the EDG 15. Each EDG does not trip on a signal generator differential relay, low-low equipment protection by-pass functions which simulates equipment protection engine oil pressure, and generator using signals which simulate the functions except for the EDG overspeed, voltage-controlled overcurrent trips, protective function trips. generator differential relay, low-low EDG equipment protection functions are engine oil pressure, and generator by-passed when an automatic start signal voltage-controlled overcurrent trips.

is received by an EDG.

16.a) Displays of EDG operating status, 16.a) Inspection for the existence or 16.a) Display of the EDG instrumentation voltage, amperage, frequency, engine retrievability in the MCR of indicating operating status, voltage, speeds, starting air pressure, and engine instrumentation displays will be arnperage, frequency, engine speeds, cooling water temperature and pressure performed. starting air pressure, and engine cooling exist in the MCR or can be retrieved water temperature and pressure exist in there. the MCR or can be retrieved there.

2.6.2 06-18-93

' SYSTEM 80+ TABLE 2.6.2-1 (Continuc_d).

ONSITE STANDHY AC POWER SOURCES Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses x .,tance Criteria 16.b) Controls exist in the MCR to start, load, 16.b) Tests will be performed using the EDG 16.t,, c'_. ', e.es in the MCR operate to

, and stop each EDG. controls in the MCR.  : rt, load and c+op each EDG.

16.c) Alarms exist in the MCR to indicate 16.c) Tests of the EDG alarms for EDG 16.c) The EDG alarms for bypasses and EDG protective function bypass and protective function bypasses and inoperability status activate in response

+ inoperability status. inoperability status will be performed to signals simulating alarm conditions.

using signals that simulate. bypass and incperability conditions.

17.a) In response to a loss of power signal, 17.a) A test of the CT will be performed 17.a) Upon receipt of a signal simulating a the CT automatically starts within 2 using a signal that simulates a loss-of- loss-o.' power, . the . CT starts in 2 minutes of receipt of a start signal and power condition, minutes or less, connects to the then automatically connects to and loads permanent non-safety bus, and loads in the permanent non-safety buses, accordance with its load sequence.

17.b) The CT car also be manually connected 17.b) Inspection of the as-built CT will be 17.b) ne CT can be connected to the Class to the Class IE busen. performed. IE buses of both Class IE Divisions.

2.6.2 06-18-93

. _ . _ _ _ _ - - . _ _ . _ . - _ - . _ . _ _ _ . . _ . . . . _ _ . _ . . _ . . _ - _ . . - _ . . . . _ . . _ _ . . - . . . . _ _ _ , _ . _ - - . _ ... a

i SYSTEM 80+" j 2.7.9 PROCESS SAMPLING SYSTEM Design Description '

l l

The Process Sampling System (PSS) collects and delivers samples from process  !

systems to sample stations for analyses. Portions of the system which form part of the ]

reactor coolant pressure boundary are safety-related. j

'Ihe PSS is located within the nuclear island structures.

The Basic Configuration of the PSS is as shown on Figure 2.7.9-1.

The ASME Code Section III Class for the PSS pressure retaining components shown on Figure 2.7.9-1 is as depicted on the Figure.

The safety related equipment shown on Figure 2.7.9-1 is qualified Seismic Category I. -

Displays of the PSS instrumentation shown on Figure 2.7.9-1 exist in the main control l room (MCR) or can be retrieved there.

4 Controls exist in the MCR to open and close those power operated valves shown on Figure 2.7.9-1. PSS alarms shown on Figure 2.7.9-1 are provided in the MCR. 1 Valves with response positions indicated on Figure 2.7.9-1 change position to that indicated on the Figure upon loss of motive power.

Inspections, Tests, Analyses, and Acceptance Criteria  ;

Table 2.7.9-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Process Sampling System.

i r

I 2.7.9 06-17-93

SYSTEM 80 o IAswe cooe secuow m CLAS5[

TO VCT (CVps)

Qm PROCESS RADIATION A MONITOR  %

[7URIbCA ON FILTER (CVCS)~,

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..q e w m ewr g cw-rwr Q +g "M g ,t FC l HOT EG l SAMPLE ,

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1. THE ASME CODE SECTION 111 CLASS 2 COMPONENTS SHOWN ARE SAFETY.RELATED

2.

• EQUIPMENT FOR WHICH PARAGRAPH NUMBER 3 OF THE " VERIFICATION FOR BASIC CONFIGURATION FOR SYSTEMS' SECTION OF THE GENERAL PROVISIONS (SECTION 1.2) APPUES FIGURE 2.7.9-1 PROCESS SAMPLING SYSTEM

SYSTEM 80+ TAHLE 2.7.9-1 PROCESS SAMPLING SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment InSDections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the PSS is 1. Inspection of the as-built PSS con- 1. For the components and equipment as shown on Figure 2.7.9-1. liguration will be conducted. shown on Figure 2.7.9-1, the as-built PSS conforms with the Basic Configuration.

2. The ASME Code Section 111 PSS 2. A pressure test will be conducted on 2. The results of the pressure test of components shown on Figure 2.7.9-1 those components of the PSS required to ASME Code Section ill components of retain their pressure boundary integrity be pressure tested by ASME Code the PSS conform with the pressure under internal pressures that will be Sectiort Ill. testing criteria in ASME Code Section experienced during service. III.

3.a) Displays of the PSS instrumentation 3.a) Inspection for the existence or 3.a) Displays of the instrumentation shown shown on Figure 2.7.9-1 exist in the retrieveability in the MCR of on Figure 2.7.9-1 exist in the MCR or MCR or can be retrieved there. instrumentation displays will be can be retrieved there.

performed.

3.b) Controls exist in the MCR to open and 3.b) Tests will be performed using the PSS 3.b) PSS controls in the MCR operate to close those power operated valves controls in the MCR. open and close those power operated shown on Figure 2.7.9-1. valves shown on Figure 2.7.9-1.

3.c) PSS alarms shown on Figure 2.7.9-1 are 3.c) Tests of the PSS alarms shown on 3.c) The PSS alanns shown on Figure provided in the MCR. Figure 2.7.9-1 will be performed using 2.7.9-1 actuate in the MCR in response signals simulating alarm conditions. to signals simulating alarm conditions.

2.7.9 06-17-93 i

SYSTEM 80+ TABLE 2.7.9-1 (Continucd)

PROCESS SAMPLING SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment inspections. Tests. Analyses Acceptance Criteria

4. Check valves shown on Figure 2.7.9-1 4. Tests will be performed to open and/or 4. Each check valve shown on Figure will open and/or close under system close check valves under preoperational 2.7.9-1 opens and/or closes.

pressure, fluid flow conditions, or differential pressure or fluid flow temperature conditions. conditions, or temperature conditions.

5. Valves with response positions indicated 5. A test of loss of motive power to these 5. These valves change position to the on Figure 2.7.9-1 change position to valves will be performed. position indicated on Figure 2.7.9-1 on that indicated on the Figure upon loss of loss of motive power.

motive power.

4 i

5 l 2.7.9 06-17-93

4 SYSTEM 80+"

2.7.16 CHEMICAL AND VOLUME CONTROL SYSTEM Design Description The Chemical and Volume Control System (CVCS) removes coolant water from the reactor coolant system (RCS), passes the coolant water through filters and ion exchangers, adds or removes soluble boron from the coolant, provides backup spray ,

water to the pressurizer, provides cooling water to the reactor coolant pump (RCP)  !

seals, collects controlled RCP seal bleedoff, provides water to the spent fuel pool and i returns water to the RCS. The CVCS is a non-safety related system except for portions of the system which form part of the reactor coolant pressure boundary, which are safety related.

The Basic Configuration of the CVCS is as shown on Figure 2.7.16-1. Components shown on the Figure are located in the nuclear island structures.

The CVCS includes pumps, valves, tanks, heat exhangers, ion exchangers, piping, instrumentation and controls.

The ASME Code Section III Class for the CVCS pressure retaining components shown on Figure 2.7.16-1 is as depicted on the figure. The safety-related equipment shown on Figure 2.7.16-1 is qualified Seismic Category I.

Displays of the CVCS instrumentation shown on Figure 2.7.16-1 exist in the Main Control Room (MCR) or can be retrieved there. Controls exist in the MCR to start and stop the charging pumps and the dedicated seal injection pump, and to open and close those power operated valves shown on Figure 2.7.16-1. CVCS alanns are provided as shown on Figure 2.7.16-1.

The dedicated seal injection pump receives Class IE power. Each ASME Code Section III Class I letdown line isolation valve is powered from a different Class 1E Division.

Valves with response positions indicated on Figure 2.7.16-1 change position to that indicated on the Figure upon loss of motive power.

The letdown line is isolated by a safety injection actuation signal (SIAS). The RCP 4

controlled bleedoff line is isolated by a containment spray actuation signal (CSAS).

I i

l 2.7.16 06-18-93

?

l l

l I

5 e

SYSTEM 80+"  ;

Interlocks are provided so that no more than one charging pump is operating at a ,

time.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.16-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the CVCS.

i P

1 I

l F

2.7.16 06-18-93

SYSTEM 80 +" iNsioE OuTsioE CONTAINMENT l CONTAINMENT

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RCGVS (SDS) = -- -- t CIV AND RECYCLE +  ;

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NOTES:

1. THE ASME CCCE! SECTON la CLASS 1 AND 2 PRESSURE RETAMING COMPONENTS SHOWN ARE SAFETYJtELATED.

& THE DEDICATED SEAL INJECTON PUMP AND THE ASME MCG SECTION pl CLASS 1 LETDOWN LINE ISOLATON VALVES RECEfvE CLASS 1E POWER.

S. * : EQUIPMENT FOR WHICM PARAGRAPH NUMBER S OF THE VERIFICATCNS FOR Basic CONRGURATION FOR SYSTEMS SECTION OF THE GENERAL PROVtSIONS (SECTON 1.2) APPLIES.

FIGURE 2.7.16-1 CHEMICAL AND VOLUME CONTROL SYSTEM

SYSTEM 80+= TABLE 2.7.16-1 CIIEMICAL AND VOLUME CONTROL SYSTEM t

Inspections. Tests. Analyses. and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the CVCS is 1. Inspection of the as-built CVCS 1. For the components and equipment as shown on Figure 2.7.16-1. configuration will be conducted. shown on Figure 2.7.16-1, the as-built CVCS conforms with the Basic Con-figuration.

2. The ASME Code Section 11I CVCS 2. A pressure test will be conducted on 2. 'Ihe results of the pressure test of components shown on Figure 2.7.16-1 those components of the CVCS required ASME Code Section 111 components of retain their pressure boundary integrity to be pressure tested by ASME Code the CVCS conform with the pressure under internal pressures that will be Section Ill. testing criteria in ASME Code Section experienced during service. 111.

3.a) Displays of CVCS instrumentation 3.a) Inspection for the existence or retriev- 3.a) Displays of the instrumentation shown shown on Figure 2.7.16-1 exist in the ability in the MCR of instrumentation on Figure 2.7.16-1 exist in the MCR or MCR or can be retrieved there. displays will be performed. can be retrieved there.

3.b) Controls exist in the MCR to start and 3.b) Tests will be performed using the CVCS 3.b) CVCS controls in the MCR operate to stop the charging pumps and the controls in the MCR. start and stop the charging pumps and dedicated seal injection pump, and to the dedicated seal injection pump, and to open and close those power operated open and close those power-operated valves shown on Figure 2.7.16-1. valves shown on Figure 2.7.16-1.

3.c) CVCS alarms shown on Figure 2.7.16-1 3.c) Tests of the CVCS alarms shown on 3.c) The CVCS alarms shown on Figure are provided as shown on the Figure. Figure 2.7.16-1 will be performed using 2.7.16-1 actuate io response to signals signals simulating alarm conditions. simulating alarm conditions.

2.7.16 06-18-93

SYSTEM 80+" TAHLE 2.7.16-1 (Continued)

CIIEMICAL AND VOLUME CONTROL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Desien Commitment Insucctions. Tests. Analyses AcceDiance Criteria 4.a) The dedicated seal injection pump 4.a) A test will be performed on the CVCS 4.a) A test signal exists at the CVCS receives Class IE power. by providing a test signal in the Class component powered from the Class IE IE Division which supplies power to the Division under test.

dedicated seal injection pump.

4.b) Each ASME Code Section 111 Class 1 4.b) Tests will be performed on the CVCS 4.b) A test signal exists only at the CVCS letdown line isolation valve is powered by providing a test signal in only one component powered from the Clcss IE from a different Class IE Division. Class IE Division at a time. Division under test.

5. Valves with response positionsindicated 5. A test ofloss of motive power to these 5. These valves change position to the on Figure 2.7.16-1 change position to valves will be performed. position indicated on Figure 2.7.16-1 on that indicated on the Figure upon loss of loss of motive power.

motive power.

6.a) he letdown line is isolated by a safety 6.a) A test will be performed using a signal 6.a) The two CVCS letdown isolation valves injection actuation signal (SIAS). simulating an SIAS. inside containment close upon receipt of a signal simulating an SIAS.

6.b) The RCP seal controlled bleedoff line is 6.b) A test will be performed using a signal 6.b) The RCP seal controlled bleedoff line isolated by a containment spray simulating a CSAS. isolation valves close upon receipt of a actuation signal (CSAS). signal simulating a CSAS.

7. Interlocks are provided so that no more 7. Tests will be performed by attempting to 7. The idle charging pump will not start than one charging pump is operating at start each charging pump from the MCR when the other pump is running.

a time, with the other pump ruring.

2.7.16 06-18-93

SYSTEM 80+" TABLE 2.7.16-1 (Continued)

CHEMICAL AND VOLUME CONTROL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

8. Motor Operated Valves (MOVs) having 8. Tests will be performed to open and/or 8. Each MOV having an active safety an active safety function will open close MOVs having an active safety function opens and/or closes.

and/or close under differential pressure function under preoperational differential or fluid flow conditions and under pressure or fluid flow conditions and temperature conditions. under temperature conditions.

9. Check valves shown on Figure 2.7.16-1 9. Tests will be performed to open and/or 9. Each check valve shown on Figure will open and/or close under system close check valves under system pre- 2.7.16-1 opens and/or closes.

pressure, fluid flow conditions, or operational pressure, fluid flow temperature conditions. conditions, or temperature conditions. ,

6 t

6 06-18-93 2.7.16 .

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I i

1 i

SYSTEM 80+" i 2.7.24 FIRE PROTECTION SYSTEM Design Description The Fire Protection System (FPS) is a non-safety related system that provides fire detection and suppression capabilities and mitigates fire propagation. The FPS consists of a water distribution system, suppression systems, a fire detection and alarm system, and portable tire cxtinguishers. I The Basic Configuration of the FPS water distribution system is as shown on Figure 2.7.24-1. Each fire protection water supply tank has a capacity of at least 300,000 i

gallons. Two fire pumps, one electric motor driven and one diesel engine driven, are provided. The electric motor driven fire pump and the diesel engine driven fire pump are separated by a three-hour fire barrier. The electric motor driven fire pump is powered from a permanent non-safety bus. A diesel fuel oil storage tank is sized to provide at least an eight hour fuel supply to the diesel engine driven fire pump. A j jockey pump is used to maintain fire protection water distribution system pressure.

Standpipe systems have piping connections to the fire protection water distribution system, isolation valves, and fire hoses. Water is supplied to the standpipe system 7 from the fire protection water distribution system.

Standpipe systems provided in the nuclear annex and in the reactor building are designed to remain operational following a design basis earthquake. The standpipe systems in the nuclear annex and in the reactor building can be supplied water from a seismically qualified backup water supply. 'Ihe backup water supply has a capacity of at least 18,000 gallons.  ;

Automatic sprinkler systems are provided for fire suppression. The sprinkler systems receive water from the fire protection water distribution system.

Manual pull stations or individual fire detectors provide fire detection capability and can be used to initiate fire alarms. Batteries supply backup power for the fire detection and alarm system.

Portable fire extinguishers are provided for fire suppression.

A plant fire hazards analysis considers potential fire hazards, determines the effects of fires on the ability to shutdown the reactor and to control the release of radioactivity to the emironment, and specifies measures for fire prevention, fire detection, fire suppression, and fire containment.

2.7.24 06-18-93 4

1-SYSTEM 80+"

Inspections, Tests, Analyses and Acceptance Criteria Table 2.7.241 specifies the inspections, tests, analyses, and associated acceptance criteria for the Fire Protection System.

l l

2.7.24 06-18-93

SYSTEM 80+*

r 3 MAKEUP _ _p WATER FIRE PROTECTION SYSTEM WATER SUPPLY TANK i

k J f

d +

ELECTRIC MOTOR DRIVEN FIRE PUMP W- TO ONSITE y 7 Q

+

BUILDING FIRE-PROTECTION SPRINKLERS AND STANDPIPES AND DIESEL MOTOR DRIVEN FIRE PUMP N

O JOCKEY

( l PUMP ,

MAKEUP _ _ _ p WATER FIRE PROTECTION SYSTEM WATER SUPPLY TANK L J FIGURE 2.7.24-1 i FIRE PROTECTION WATER DISTRIBUTION-SYSTEM

SYSTEM 80+= TABLE 2.7.24-1 FIRE PROTECTION SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Desien Commitment inspections. Tests. Analyses Acceptance Criteria

1. The Basic Configuration of the FPS 1. Inspection of the as-built FPS water 1. For the components and equiparnt water distribution system is as shown on distribution system configuration will be shown on Figure 2.7.24-1, the as-built Figure 2.7.24-1. conducted. FPS water distribution system conforms with the Basic Configuration.

2. Each fire protection water supply tank 2. Inspections of the as-built fire protection 2. Each fire protection water supply tank has a capacity of at least 300,000 water supply tanks will be performed. has a capacity of at least 300,000 gallons. gallons.

3. The electric motor driven fire pump and 3. Inspection of the as-built fire barrier 3. The electric motor driven fire pump and the diesel engine driven fire pump are will be performed. the diesel engine driven fire pump are separated by a three-hour fire barrier. separated by a three-hour fire barrier.

4. The electric motor driven fire pump is 4. Tests will be performed on the FPS by 4. Within the FPS, a test signal exists at powered from a permanent non-safety providing a test signal in the permanent the equipment powered by the bus. non-safety bus. permanent non-safety bus under test.

5. A diesel fuel oil storage tank is sized to 5. A test of the fuel consumption of the 5. The diesel fuel oil storage tank has at provide at least an eight hour fuel diesel engine driven fire pump will be least an eight hour fuel supply for the supply to the diesel engine driven fire performed. An inspection of the fuel diesel engine driven fire pump.

pump. supply tank will be performed. The fuel supply capacity will be determined.

6. *The backup water supply to the 6. An inspection of the as-built backup 6. The backup water supply has a capacity standpipe system in the nuclear annex water supply will be performed. of at least 18,000 gallons.

and the reactor building has a capacity of at least 18,000 gallons.

2.7.24 06-18-93

SYSTEM 80+ TAHLE 2.7.24-1 (Continued)

FIRE PROTECTION SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment inspections. Tests. Analyses Acceptance Criteria

7. Batteries supply backup power for the 7. A test of the fire detection and alarm 7. The fire detection and alarm system is fire detection and alarm system. system will be conducted under a provided battery-supplied backup power.

simulated loss of power.

8. A plant fire hazards analysis considers 8. A fire hazards analysis will be 8. A fire hazards analysis exists and potential fire hazards, determines the performed. considers potential fire hazards, effects of fires on the ability to determmes the effects of fires on the shutdown the reactor and to control the ability to shutdown the reactor and to release of radioactivity to the contain the release of radioactivity to the environment, and specifies n,easures for environment, and specifies measures for fire prevention, fire detection, fire fire prevention, fire detection, fire suppression, and fire containment. suppression, and fire containment.

2.7.24 OG18-93

i

[

SYSTEM 80+"

2.7.25 COMMUNICATIONS SYSTEMS Design Description The Communications Systems are non-safety-related systems that provide onsite ,

communications capability and means to communicate with offsite specified  ;

participating entities. The Communications Systems consist of a Portable Wireless Communication System, a Private Automatic Business Exchange (PABX) Telephone System, a Public Address (PA) System, a Sound-Powered Telephone System, and an  ;

Offsite Communications System.

l The Portable Wireless Communication System provides communications capability among control room operators, equipment operators, and maintenance technicians for routine and emergency operations.

The Private Automatic Business Exchange (PABX) Telephone System and the Public Address (PA) System are provided as alternate means of communications. The PABX Telephone System provides intraplant communications and access to offsite telephone systems. The PA System provides a means to alert plant personnel through audible speakers located throughout the plant.

The intraplant Sound-Powered Telephone System uses phone jacks which can be patched together to establish communications between areas of the plant where maintenance, refueling, or shutdown operations are conducted.

In addition to the PABX interface with the offsite telephone system, direct offsite communications, independent of the PABX, are provided to the plant and support facilities. The direct offsite emergency telephones are identified distinctly from the PABX telephones. The emergency telephones provide links with the Nuclear Regulatory Commission (NRC) and specified participating local and state agencies.

A security radio system and a crisis management radio system are provided for communication between specified participating entities.

Imss of electrical power to any of the Communications Systems does not affect the operability of the remaining Communications Systems.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.25-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Communications Systems.

2.7.25 06-16-93

SYSTEM 80+" TABLE 2.7.25-1 COMMUNICATIONS SYSTEMS Inspections. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. He Portable Wireless Communication 1. Tests - of . the Portable Wireless 1. Voice transmission and reception are System provides contmunications cap- Contmunication System will be. accomplished.

ability among control room operators, performed.

equipment operabrs, and maintenance technicians for routine and emergency operations.

2. The PABX Telephone System provides 2. Tests of the PABX Telephone System 2. Voice transmission and reception intraplant communications and access to will be performed. between plant termmals are accom-offsite telephone systems. plished. Voice transmission and  !

reception between onsite terminals and the offsite telephone systems are accomplished.

3. The PA System provides a means to 3. Tests of the PA System will be 3. Voice transmission and reception are alert plant personnel through audible performed. accomplished.

speakers located throughout the plant.

~

4. He intraplant Sound-Powered Tele- 4. Tests of the intraplant Sound-Powered 4. Voice transmission and reception are phone System uses phone jacks which Telephone System will be performed. accomplished.

can be patched together to establish conununications between areas of the plant where maintenance, refueling, or

, shutdown operations are conducted.

2.7.25 . 1- 06-16-93 .

b

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SYSTEM 80+ TAl;LE 2.7.25-1 (Continued)

COMMUNICATIONS SYSTEMS inspections Tests Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria  !

5.a) In addition to the PABX interface with 5.a) Tests of the offsite telephone system will 5.a) Voice transmission and reception are the offsite telephone system. direct be performed. accomplished to the NCR and specified t offsite communications, independent of participating local and state agencies.

the PABX, are provided to the plant and support facilities. The emergency tele-phones provide links with the NRC, and specified participating local and state agencies.

5.b) The direct offsite emergency telephones 5.b) Inspection of the offsite emergency 5.b) The direct offsite emergency telephones are identified distinctly from the PABX telephones will be performed. are color coded to distinguish them from Telephone System. the PABX telephones.

6. A security radio system and a crisis 6. Tests of the recurity radio system and 6. Two way communication between speci-numagement radio system are provided the crisis management radio system will fled participating entities is demon-to provide communications between be performed. strated.

specified participating entities.

i

7. less of electrical power to any of the 7. Tests for operability of the Communi- 7. Loss of power to any of the Communi-Communications Systems does not affect cations Systems will be performed with cations Systems does not disrupt the the operability of the remaining Com- actual or simulated loss of electrical voice transmission and reception cap-munications Systems. power conditions. abilities of the remaining Communi-cation Systems.

2.7.25 06-16-93

- , . . - - , ~ e-r... #%.. es%%e w. -e-a ......-w..- .

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r i

SYSTEM 80+" ,

2.7.26 LIGHTING SYSTEM Design Description The Lighting System is a non-safety-related system that is used to provide illumination ,

at locations in the plant and on the plant site. The Lighting System has a normal  :

lighting system, a security lighting system, and an emergency lighting system.

ne normal lighting system provides general illumination at locations in the plant.

The security lighting system provides illumination in isolation zones and outdoor areas I within the plant protected perimeter. The security lighting system is powered from j the permanent non-safety buses. I l

l The emergency lighting system provides illumination in the vital areas that include the l main control room (MCR), the technical support center, the operations support l center, the remote shutdown room, and the stairway which provides access from the MCR to the remote shutdown room.

Emergency lighting in the MCR is provided such that at least two strings oflighting j fixtures are powered from different Class 1E Divisions. He emergency lighting in the  ;

MCR maintains minimum illumination levels in the MCR during emergency  !

conditions including station blackout. The emergency lighting installations which j serve the MCR are designed to remain operational following a design basis j earthquake.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7.26-1 specifies the inspections, tests, analyses and associated acceptance criteria for the Lighting System.

1 1

1 2.7.26 06-18-93

SYSTEM 80+ TABLE 2.7.26-1 LIGIITING SYSTEM Inspections. Tests. Analyses and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The security lighting system is powered I. Tests will be performed on the security 1. Within the security lighting system, a from the permanent non-safety buses. lighting by providing a test signal in the test signal exists at the equipment permanent non-safety buses. powered by the permanent non-safety bus under test.

2. The emergency lighting system provides 2. An inspection of the MCR, the technical 2. Emergency lighting is installed in the i!!umination in the vital areas that support center, the operations support (MCR), the technical support center, the include the MCR, the technical support center, the remote shutdown room, and operations support center, the remote center, the operations support center, the stairway which provides access from shutdown room, and the stairway which the remote shutdown room, and the the MCR to the remote shutdown room provides access from the MCR to the stairway which provides access from the will be performed. remote shutdown room and provides MCR to the remote shutdown room. illumination levels greater than or equal to 10 foot-candles in the MCR, technics 1 support center, operations support center, the remote shutdown panel room and the stairway which provides access from the MCR to the remote shutdown tuom.

3. Emergency lighting in the MCR is 3. Tests will be performed on the 3. Within the MCR emergency lighting provided such that at least two strings of emergency lighting system in the MCR system, a test signal exists only at the lighting fixtures are powered from by providing a test signal in only one equipment powered fmm the Class IE different Class IE Divisions. Class IE Division at a time. Division under test.

4. The emergency lighting in the MCR 4. Tests of the emergency lighting system 4. Under simulated station blackout maintains minimum illumination levels will be performed under simulated conditions, the emergency lighting in the MCR during emergency station blackout conditions. system in the MCR maintains conditions including station blackout. illumination levels greater than or equal to 10 foot-candles, i

i i 2.7.26 06-18-93 i

i

A SYSTEM 80+"

2,9.1 LIQUID WASTE MANAGEMENT SYSTEM Design Description The Liquid Waste Management System (LWMS) is a non-safety-related system which is used to collect, segregate, store, process, sample, and monitor radioactive liquid waste.

The LWMS is located in the radwaste building.

The Basic Configuration of the LWMS is as shown on Figure 2.9.1-1.

The LWMS has four subsystems which process radioactive or potentially radioactive liquid waste. These four subsystems segregate liquid waste into high level waste, low level waste, laundry and hot shower / chemical waste, and the containment cooler condensate waste.

The high level waste subsystem has filters, demineralizers, provisions for batch sampling, and piping for recirculation of liquid waste for further processing.

The low level waste subsystem has filters, demineralizers, provisions for batch sampling, and piping for recirculation of liquid waste for further processing. i l

The laundry and hot shower / chemical waste subsystem has filters, deminerahzers, -

provisions for batch sampling, and piping for transfer of laundry and hot j shower / chemical wastes to the low level waste subsystem for further processing.

The containment cooler condensate subsystem has tanks to collect containment cooler l condensate. He discharge from the tanks is monitored for radioactivity. Although not normally radioactive, this discharge can be diverted to the low level waste subsystem.

The LWMS subsystems have collection and storage capacity to process waste volumes expected during normal operation and from anticipated operational occurrences.

Displays of the LWMS instrumentation shown on Figure 2.9.1-1 exist in the main control room (MCR) or can be retrieved there.

Controls exist in the MCR to open and close the power operated valve shown on Figure 2.9.1-1.

I 2.9.1 06-18-93

SYSTEM 80+"

The LWMS has means to monitor radioactivity levels in the processed liquid waste prior to release. The radioactivity monitor provides a signal to terminate LWMS discharge when a specified radioactivity level is reached.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.9.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Liquid Waste Management System.

l f

i i

l 2.9.1 06-18-93 1

l

-SYSTEM 80 +*  ;

l"suCLEAR Ab"NEX l U EQUIPMENT -> r l DRAINS l t

I RADWAS5E BLDG l EQUIPMENT ->

l DRAINS l HIGH LEVEL  !

+ WASTE  :

SUBSYSTEM l_ TURBINE BLDG.____l l EQUIPMENT (IF RADIOACTIVE)DRAINS l _

l SG DRAINS l SG EQUIPMENT -*

I DRAINS l 2

~

- l REACTOR CAVITY l

-+

l & CONTAINMENT SUMP l

! NUCLEAR ANNEX  !

g FLOOR DRAINS l

~ ~~~~

+ WASTE  ;

SUBSYSTEM HIGH RADIOACTIVITY LEVEL g FUEL BLDG.---g i

g FLOOR DRAINS g"""

I TURBINE BLDG. I FLOOR DRAINS - ->

' l JF R,A,DIOACTIVE) _!

i

TO DISCHARGE FC  :

l LAUNDRY i_ _ "^_'"*_ _l l

,_ REGULATED SHOP , i DRAINS l j LAUNDRY AND HOT I

"" """ ~~ """ ~

.+ SHOWER / CHEMICAL -

= 1 WASTE l PERSONNEL l SUBSYSTEM ,

DECON. -> .l l SHOWERS l

~

ICASK CLEANING _____l )'

lMISC.

DETERGENTEQUIP. DRAINS SAMPLE l- > l l

1

- l CHEMICAL DECON. l l W

I _ASTES,L_AB D_RAI_NS _

SAM PLES,l"""

• l i

CONTAINMENT i l., CONTAINMENT l COOLER  !

COOLER & CONDENSATE O l CONDENSATE DRAINS l SUBSYSTEM i FIGURE 2.9.1-1  :

LIQUID WASTE MANAGEMENT SYSTEM i 1

1

SYSTEM 80+" TAHLE 2.9.1-1 LIOUID WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acccotanee Criteria

1. The Basic Configuration of the 1. Inspection of the as-built LWMS 1. For the components and equipment equipment drain waste subsystem is as configuration will be cenducted. shown on Figure 2.9.1-1, the as-built shown on Figure 2.9.1-1. LWMS conforms with the Basic Configuration.

2. He LWMS subsystems have collection 2. Analysis of the as-built LWMS 2. An analysis exists which concludes the and storage capacity to process waste subsystems' processing capability will LWMS subsystems have collection and volumes expected during normal be performed. storage capacity to pmcess waste operation and from anticipated volumes expected during normal operational occurrences. operation and from anticipated operational occurrences.

3. Displays of the LWMS instrumentation 3. Inspection for the existence or 3. Displays of the instrumentation shown shown on Figure 2.9.1-1, exist in the retrievability in the MCR of on Figure 2.9.1-1, exist in the MCR or MCR or can be retrieved there. instrumentation displays will be can be retrieved there.

performed.

4. Controls exist in the MCR to open and 4. Tests will be performed using the 4. LWMS controls in the MCR operate to close the power operated valve shown LWMS controls in the MCR. open and close those power operated l on Figure 2.9.1-1. valve shown on Figure 2.9.1-1.

5. De radioactivity monitor provides a 5. Test of the as-built LWMS discharge 5. LWMS discharge is terminated when the l

signal to terminate LWMS discharge controls will be performed using a simulated radioactivity level in the when a specified radioactivity level is signal which simulates radioactivity discharge waste line reaches a specified reached. levels. limit.

2.9.1 06-18-93

SYSTEM 80+"

2.9.2 GASEOUS WASTE MANAGEMENT SYSTEM Design Description P

The Gaseous Waste Management System (GWMS) is a non-safety-related system which is used to collect, store, process, sample, and monitor radioactive gaseous waste.

The GWMS is located in the nuclear island structures.

The Basic Co<guration of the GWMS is as shown on Figure 2.9.2-1.

The GWMS processing unit has a cooler condenser, a charcoal guard bed, and charcoal adsorbers. A gas analyzer is provided to monitor the concentration of oxygen in the GV/MS.

Displays of the GWMS instrumentation shown on Figure 2.9.2-1 exist in the main control room (MCR) or can be retrieved there.

Controls exist in the MCR to open and close the power operated valve shown on Figure 2.9.2-1.

The GWMS provides a means to monitor radioactivity leveis in the processed gaseous ,

waste prior to release through the unit vent. The radioactivity monitor provides a signal to terminate GWMS discharge when a specified radioactivity level is reached.

Inspections, Tests, Analyses, and Acceptance Criteria Table 2.9.2-1 specifies the inspections, tests, and analyses, and associated acceptance criteria for the Gaseous Waste Management System.

2.9.2 06-18-93 ,

SYSTEM 80 +*

HIGH RADIOACTIVITY LEVEL i _. _ _. _ _ _ _ I ^

GAS STRIPPER I

(CVCS) l~ "" > l~ NITROGEN"" ~ ~ 1 -

i gas 1  :

i_. _ _ _ __ _ _i

~~

I EQUIPMENT DRAIN I N 1 TANK I--( b }- -

l- _ >

I GASEOUS WASTE h g----

TO UNIT l

~

• PROCESSING DhFC I VENT l

I VOLU51E CONTROL ~I UNIT _ __

i TANK l--> I g _ _ _(CVCS) y__

l~ REACTOR DRAUN I i_ DRAINS TO l LwMs l I TANK l- - > ~~~~

I _(C,VC,S)_ _ ,, i l

FIGURE 2.9.2-1 GASEOUS WASTE MANAGEMENT SYSTEM

SYSTEM 80+ TABLE 2.9.2-1 ,

GASEOUS WASTE MANAGEMENT SYSTEM InSDeClions. Tests. Analyses. and Acceptance Criteria Design Commitment Inspections. Tests. Analyses Acceptar.cc Criteria

1. He Dasic Configuration of the GWMS 1. Inspection of the as-built GWMS 1. For the components and equipment  !

is as shown on Figure 2.9.2-1. configuratios will be conducted. shown on Figure 2.9.2-1, the as-built GWMS conforms with the Basic Configuration. ,

2. Displays of the GWMS instrumentation 2. Inspection for the existence or 2. Displays of the instrumentation shown shown on Figure 2.9.2-1 exist in the retrievability in the MCR of on Figure 2.9.2-1 exist in the MCR or J MCR or can be retrieved there. instrumentation displays will be can be retrieved there.

performed.

3. Controls exist in the MCR to open and 3. Tests will be performed using the 3. GWMS controls in the MCR operate to close the power operated valve shown GWMS controls in the MCR. open and close the power operated valve on Figure 2.9.2-1. shown on Figure 2.9.2-1.

4. He radioactivity monitor provides a 4. Test of the as-built GWMS discharge 4. GWMS discharge is terminated when signal to terminal GWMS discharge controls will be performed using a the simulated radioactivity level in the when a specified radioactivity level is signal which simulates radioactivity discharge waste line reaches a specified reached. levels. limit.

4 I

2.9.2 06-18-93

i SYS)EM 80+"'

2.9.3 SOLID WASTE MANAGEMENT SYSTEM f Design Description i

The Solid Waste Management System (SWMS) is a non-safety-related system which 1 is used to collect, segregate, decontaminate, process, sample, and store radioactive i solid waste.

The SWMS is located in the radwaste building.

The Basic Configuration of the SdMS is as shown on Figure 2.93-1.

Solid waste is segregated into the following:

High activity and low activity wetted waste, e.g. spent ion exchanger resin and !'

spent filter assemblies; and, i

Compactible and non-cornpactible dry solid waste, e.g. plastic sheeting, clothing, or metal tools. l The high activity and low aethity spent resin processing units have collection and i storage capacity to process waste volumes generated during normal operation and i from anticipated operational occurrences. These subsystems can process waste by dewatering in the shipping container.

The dry solid material subsystems have provisions for sorting of wastes, compaction of compactible waste and placement in shipping containers, and for either decontamination or direct placement of non-compactible waste into shipping ccmtainers.

Inspections, Tests, Analyses, and Acceptance Critola Table 2.93-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Solid Waste Management System.

2.9.3 SYSTEM 80 +*

LOW ACTIVITY __

,_I SPENT RESIN l DEMINERAUZER l l ,,SPENT FILTER SUBSYSTEM (SWMS) g WATER MAKEUP SYSTEM I I ASSEMBUES I

?lg l

l l ~~ FUEL ~ ~POOL ~~~l lf I LOW LEVEL

-% 1 SORTING l

! ION EXCHANGER l-

• PACKAGING

. __ . HIGH ACTIVITY SPENT l& HANDLING '

- RESIN SUBSYSTEM --

AREA A l- - - - -l l l l PREHOLDUP ION I RADWASTE BLDG p

y EXCHANGER g+ 1 VENTil.ATION Nd g

> < '-----1 (CVCS)

Hl

. _ _ _ _ _ _ . 8 - SYSTEM

- - 8 -----

~

if l fJ L l~PU5FICAT10N N l niGH DEWATERED g EXHANGER p+ l INTEGRITY l WASTE (CVCS) CONTAINER PROCESSING I I AREA

- __. _ _ _ _ _i lf E TRUCK MOUNTED RESIN DEWATERING JL

,.T if' CONTAINER

---

O PUMP r _' g lf l~ LAUNDRY, HOT l SHOWER TANK I

I ION EXCHANGER I~~

! i (LWM.S,9 I FLOOR DRAIN TANK I g ION EXCHANGER g_ + 37 LOW ACTIVITY (LWMS) y SPENT RESIN y SUBSYSTEM

, EQUIPMENTWASTE ,

l l TANK ION EXCHANGER l= --g>-

FIGURE 2.9.3-1

. _ __ t*. "_S) _ _.

SOLID WASTE MANAGEMENT SYSTEM i - _ _ . .

SYSTEM 80+ TABLE 2.93-1 SOLID WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

1. He Basic Configuration of the SWMS 1. Inspection of the as-built SWMS 1. For the components and equipment is as shown on Figure 2.9.3-1. configuration will be conducted. shown on Figure 2.9.3-1, the as-built SWMS confonns with the Basic Configuration.

2. He high activity and low activity spent 2. Analysis of the as-built spent resin 2. An analysis exists which concludes that resin processing units have collection subsystems' processing capability will the spent resin processing units have and storage capacity to process waste be performed. collection and storage capacity to volumes expected during normal process waste volumes expected during operation and from anticipated normal operation and from anticipated operational occurrences. operational occurrences.

2.9.3 _ _ -. . _ _ - _ _ _ - _ _ - _ - - _ - _ _ - _ _ _ - _ _ - _ _ _ _ _ _ ~ . _ - _ . . _ _ _ _ . . _ .

I i

SYSTEM 80+"  ;

2.10 TECHNICAL SUPPORT CENTER  ;

Design Description -

i The Technical Support Center (TSC) performs a non-safety related function and is located adjacent to the main control room (MCR) in the nuclear annex. The TSC provides facilities for management and technical support to plant operations during emergency conditions. ,

The TSC is located less than or equal to two minutes walking time from the MCR.

The TSC has floor space of at least 75 square feet per person for a minimum of 25 persons. ,

The TSC has radiation detection equipment for monitoring radiation levels when the TSC is in use.

i The TSC has means for voice communication to the MCR, to on-site emergency support facilities, and to off-site via dedicated or commercial telephone networks.'

J Displays of the information from the discrete indication and alarm system (DIAS) and the data processing system (DPS) exist in the TSC or can be retrieved there.2 Inspections, Tests, Analyses and Acceptance Criteria l

Table 2.10-1 specifies the inspections, tests, analysis, and associated acceptance criteria for the Technical Support Center.

l 4

l

' Communication Systems are addressed in Section 2.7.25.

l 2 '

Display information from the DIAS and DPS is addressed in Section 2.53.

2.10 06-18-93

SYSTEM 80+" TAHLE 2.10-1 TECHNICAL SUPPORT CENTER InSDections. Tests. Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 1.a) The TSC is located less than or equal to 1.a) A test of walking time from the TSC to 1.a) The TSC can be reached in less than or two minutes walking time from the the MCR will be performed. equal to two minutes walking time from MCR. the MCR.

1.b) The TSC has floor space of at least 75 1.b) Inspection of the TSC will be 1.b) Floor space of at least 1875 sq. ft. is square feet per person for a minimum of performed. provided in the TSC.

25 persons.

1.c) He TSC has radiation detection 1.c) An inspection of the radioactivity 1.c) Radiation detection equipment to equipment for monitoring radiation detection equipment in the TSC will be monitor radiation levels is available in levels when the TSC is in use. performed. the TSC.

1.d) De TSC has means for voice 1.d) An inspection of the TSC will be 1.d) Communications equipment is installed.

communications to the MCR, to on-site perfsrmed.

emergency support facilities, and to off-site via dedication or commercial telephone networks.

2. Displays of information from the DIAS 2. Inspection for the existence or 2. Displays of information from the DIAS and the DPS exist in the TSC or can be retrievability in the TSC of the and the DPS exist in the TSC or can be retrieved there. information from the DIAS and the DPS retrieved there.

will be performed.

2.10 06-18-93

SYSTEM 80+"

3.1 PIPING DESIGN Design Description Piping classified as Seismic Category I is required to withstand the effects of a safe shutdown earthquake (SSE), maintain dimensional stability, and remain functional Seismic Category I piping, structures, systems and components assure: (1) the integrity of the reactor coolant pressure boundary, and (2) the capability to shut down 1

the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Piping classified as Seismic Category II is piping that is not required to function during or after a seismic event, but whose failure or interaction with Category I equipment could prevent a safety system from achieving its safety function or result in an incapacitating injury to occupants of the control room.

Seismic Category I piping and Category II piping are designed to meet the requirements of the ASME Code,Section III.

Applicable piping loads due to pressure, gravity, thermal expansion, seismic excitation, wind, tornado, fluid transients, thermal stratification, missiles, and postulated pipe breaks are considered in the piping analyses.

The as-built ASME Code Section III piping will be reconciled with the piping design l requirements described herein.

Piping system design considers the effects of crosion, corrosion, waterhammer and steam hammer. Piping system supports for Seismic Category I and II piping systems are designed to meet the requirements of the ASME Code Section III, Subsection NF. Pipe loads applied to attached equipment are shown to be less than the equipment allowable loads.

Piping systems classified as ASME Code Section III Class 1,2, or 3 are designed to maintain dimensional stability and functional integrity under design loadings expected to be experienced during a 60 year design life.

Design of piping systems provides for clearances between adjacent piping, components, and other structures when the piping moves due to design static, dynamic, and thermal loadings.

3.1 06-17-93

c q

1 l

SYSTEM 80+"' ,

l The following piping systems are designed to meet leak-before-break (LBB) criteria:

Reactor coolant system hot leg piping, reactor coolant pump (RCP) suction ,

piping and RCP discharge piping, ,

Surge line, Main steam lines inside containment from the steam generator to the first anchor on the main steam lines, Shutdown cooling lines inside containment from the reactor coolant system to the first anchor on the shutdown cooling lines, and Direct vessel injection lines inside containment from the reactor vessel to the first anchor on the direct vessel injection lines LBB acceptance criteria are established and LBB evaluations are performed for each piping system designed to meet LBB criteria. For each piping system qualified for LBB, the as-built piping and materials will be reconciled with the bases for the LBB acceptance criteria.

Structures, components, equipment and systems required for safe shutdown are protected from the dynamic effects of postulated pipe breaks not eliminated by LBB.

Design of features which protect these items consider, as applicable, pipe whip, water spray, jet impingement, flooding, compartment pressurization, and emironmental conditions in the area where the piping is located.

Inspections, Tests, Analyses and Acceptance Criteria Table 3.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Piping Design.

l l

3.1 06-17-93

d SYSTEM 30+" TAHLE 3.1-1 PIPING DESIGN Inspections. Tests. Analyses, and Acceptance Criteria Desien Commitment Inspections. Tests. Analyses Acceptance Criteria

1. The as-built piping is reconciled with 1. Inspection will be performed of the 1. The as-built piping is reconciled with the as-designed piping configurations. piping system to confirm as4esigned the pipag design requirements described conditions. The following will be in the piping design description.

verified: pipe routing, support and equipment types and locations, required clearances, and orientation of valve operators. I l

2. Piping systems classified as ASME 2. Inspation for the existence of ASME 2. ASME design reports for piping systems I Cale Section III Class I, 2, or 3 are design reports will be performed. classified as ASME Code Section 111 designed to maintain dimensional Class 1,2, or 3 exist and conclude that stability and functional integrity under the structural integrity requirements of design loadings expected to be the ASME Code are met.

experienced during a 60-year design life.

3. For each piping system qualified for 3. For each piping system qualified for 3. A LBB evaluation report documents that LBB, the as-built piping and materials LBB, an inspection of the LBB le it-before-break acceptance criteria are will be reconciled with the bases for the evaluation report will be performed. met by the as-built piping and piping LEB acceptance criteria. materials.

4 Structures, components, equipment and 4. For piping systems with postulated pipe 4. A pipe break analysis report exists and systems required for safe shutdown are breaks, an inspection of the pipe break concludes that structures, systems, and protected from the dynamic effects of report willbe performed. An inspection components classified as ASME Code postulated pipe breaks not eliminated by of the as-built high energy pipe break Section til Class I, 2, or 3 remain LBB. mitigation features will be performed. functional after postulated pipe breaks.

3.1 06-17-93

_ _ .