ML17334A523

Facility Conceptual Design Description for Technical Support Ctr & Emergency Operations Facility.

ML17334A523
Person / Time
Site:	Cook
Issue date:	09/15/1982
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
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ML17334A522	List: ML17334A521 ML17334A523
References
NUDOCS 8404030229
Download: ML17334A523 (106)

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INDIANA&MICHIGAN EXZCTRIC CCMPMY CONALD C.QXK NKXZRR PLANT KE THE TSZBGCAG SUPPORT CENTER MD THE KF.ATTAGBKÃT K)AEP:NRC: 0533A Revised: September 15, 1982 and incorporated as Attachment 2 to AEP: NRC: 0531E'hanges are indicated by a bar in the right-hand margin.'Ihis dccun nt contains iafornation pxcpxietazy to Westinctmuse Elect~Co~zation and An~can Elec"~Power Sezv'ce Cozpozaticn; it is suhnitted in confidence ard is to he used sole+for Nw purpose for which it Is fuz2LLshed

~dccQm'.?It and such infozIGaticn Ls not to he zepzoduced, tzananitted, disclosed or used at2~vtse in whole or in par" 8404030229 820929 PDR ADOCK 050003l5 F PDR gESTtr<grOUSE PROPRIETARY CLASS 2'his.document contains material that is proprietary to the Mestinghouse Electric'orporation.

The proprietary information has been marked by means of brackets.The basis for marking the material proprietary is identified by marginal notes referring tb the standards in Section 8 of the affidavit of R.A.Miesemann of record"In the Matter of Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors (Oocket No.RH-50-1)" at transcript pages 3706 through 3710 (February 24, 1972).Oue to the proprietary nature of the material contained in this report: which.was obtained at considerabIe Mestinghouse expense and the release-of which would seriously affect our competitive position, we request this information to be withheld from public disclosure in accordance with the Rules of Practice, 10 CFR 2.790, and that the information pr e-sented therein be safeguarded in accordance with 10 CFR 2.903.Me believe that-withholding this information will not adversely affect the public.interest.This information is for your internal-use only and should not be released to persons or organizations outside the Oirec.orate of Regula-tion and the ACRS without prior approval of Westinghouse Electric Corporation.

Should it become necessary to release this information to.such persons as part of the review procedure, please contact Mesting-house Electric Corporation and they will make the necessary arrangements required to protect their proprietary interests.

4RRl A Section 1.1.1 1.1.2 1.1.3 1.1.4 1.2 Title Zntzxduction System Ful~ions'Zechnical Support Center Safety Parm~xs Display System Nuclear Data Link Bypass 6 Znoperable Status Zr~tion.System Eb~rt Basis~cC AEP-1 AEP-1 AEP 1 AEP-2 AEP-2 AEP-3 2.2.1 2.2 2.3 2.3.1 2.3;2 2.3.3~Data Acquisition a Display System Cat@uter System System Ehta Display System Cnsite Technical Support Center Contxol Hocm Btarger~Cpezating Facilities 3.3e1 3.2 3.3 Cnsite Technical support Center Desian Basis Znput Detexminatian OTSC Ccexator Zntex ace AEP-9 AEP-9 AEP-10 AEP-11 4 4.1 4.2 4.3 Safety Pazaratexs Display System Purpose Znput Detemunation Man-<~hixm Zntexf ace AEP-30 AEP-30 AEP-30 AEP-33 5.571 5.2 5.3 Bypass&Jr~able Status Zndication System Purpose ManW~>e Zntexf ace AEP-47 AEP-47 AEP 47 AEP-47 6.

Section Title 7.7.1 7.1.1 7.1.2 TSC~Supply Systems~to the TSC Catguter~UPS System Cons~Ra of Pm'upply AEP-56 AEP-56 AEP-56 AEP-56 7.2 Pamr to the TSC Complex AEP-57 8.8.1 8.1.1 8.1.2 8.1.3 8.1.4 8.1.5 8.2 8.2.1 8.2.2 8.2.3 8.3 Task Functions Perfoznad by EnLLviduals in the TSC.Technical Support Yanageaant

~zt Emergency Functions Pexfozmed in the TSC/ECP for each Erargency Class.Chusual Event Alert Site and Genial Bmzgehcy Functions of Individuals Reporting to the ECF.AEP-58 AEP-58 AEP-58 AEP-58 AEP 58 AEP-59 AEP-59 AEP 59 AEP-59 AEP-60 AEP-61 AEP-62 9.9.1 9.2 9.3 TSC Record and Data Availabil'ty AEP-63 Controlled Plant Specific Beferer~i<wterial AEP-63 Chca~lled Enfozma~and Tec.'nical AEP-64 Referer~Ywtexial.Other Mta, Records, arZ Znfonraticn

~5 1.ZBZEGDKZICH 1 1 SYST124 FCKTECNS: The D.C.Cook Plant Technical~xt Center Data System is being developed and designed using the guidelines of NUB'696 to pzovide the plant cpexating and technical~xt pezsannel with tM pm~nt plant information to facilitate the end~nay response to an accident.'Ibis System, which utilizes the-Westinc~se P2500 TSC Can@uter Systans, can also be used duxing nozmal plant agezaticn for ather fhrctians such as-plant pezfonmxa analysis, pezsonnel Dmin~etc.system cansists af~similar caagutezized data acquisition, pzccessing and display systems, ere for each D.C.Cook Unit.The=four navar functions pzavided by this ccmputer system are: 1.1.1 TZGKXCAL SGPPORC CENTER (TSC): The ccnguter system will receive, stoze, prccess and display on color~tmanix~and/or cn hard-copy teaninals the real time data acquixed fxcm vaxious plant syst~.Pre-trip and post-~data are also collected ard can be pzocessed and displayed by the cancuter.This system will facilitate the assessnant af ttm plant's condition by p1ant operating ard technical smpoxt cexsonrml.

The data displays af th Te&nical~xt Cmzter fur~ion will pzenride suf icient infozmation to deterrnirw:

'~-Plant steady state cgamting-canditians.

prior ta Nm unit trip-Transient conditions pxcducing Nm initiating event and system 1x8zavior duz~the~e af the accident.-Pxesent conditions af the plant.The TSC data dzsp1ay systan may he used for.-Reviewing the accident secnzence..

-Detezznizuz@

apprcpxiate mitigating actions.-Evaluating the e'xtmzt af any damage.-Detezznizzizg plant status during recovery cgexaticm.

function will he desc~R in details in Section 3.0 1 1.2~M~FETY STATIC DISPLAY (PSSD): This PSSD systen was designed in accordance with the guidelines for the Safety Paxam te Disp1ay System (SOS)of NGREG 0696.This PSSD system,~displays the safety status af Nm plant in a faxmat that.can he easily xeax~ed hy+~contxol roam operators, will help the operators to detect any ahnoznnal ccnditian in a~ly tnanmr.Pdditiar~l features af this PSSD system will he1p the operators and technical support personnel to chtaiz>detailed information an&~safety systems af Nm plant.Detailed descriptions af this systan are plaided in Secticn 4.1 1 3 NXZZAR DATA LINK (NDL)The TSC cartcuter system has a huilt-in aff-site data txazmnissian capahili.ty which can he used for izztexfacing with a future Nuclear Data Link (NDL)Sub-Syst m.l1 AZP 2

1.1.4 am'rmeZWBrZ

STATta INDICATE Swam (BISI): The BISX system provides the operators and technical support personnel with a clear indicatian af.the availability oC Nu plant safety systems (ESF Systems).Detailed descrq~ns of this system aze provided in Section 5.1.2 REPCBT BASIS: This report is~on the proprietary Westingbmxse KRP Hegort 9725"Westinghouse Technical Support Camlex," which was submitted to the HRC.'-Appropriate mxiifiaatians wexe ttede to reflect the specific design of D.C.Ccak M.ts 1 aeR 2.

2.THE DOZE ACQUISITZCH

&DISPEL SYSTEM 2.1 GSE COMP~SYSTEM: F~2.1 shows the canguter system haxdwaxe for each Ccok Unit.Multiple 16-bit high speed minicomputer and ttenaxy devices are used to process plant data, generate displays and pexfoxm other man~chine interface functions.

The system is configured in a fault~tolerant chsign.Zf a cantxal processing unit (CPU)or a~rtion of aamxy fails, the system will automatically reconf'uxe itself to perform its chsignated functions.

2.2 ZNPOZ

SYSTEM Figure 2.2 shows Nn schematic diagram for the TSC computer System.Input signals frcm the contxol xccm ard other plant locations are taken to the xemote Input/Output (I/O)cabinets.Signal isolation is provided in the I/O cabinets so that no failure on the output side of the I/O cabinets will affect the input signals.In addition to J~se isolators, all signals conung from the safety systara are taken after the existing amlified isolators on these syst~.'Ihe input signals, after going+~gh the isolators, will be converts to bina~information on the i~cards and then axe rultiplexed to the computer.Each analog signal channel has its cwn Analog/Digital Conver~, thus providing a high degree of reliability for the input system.

C, Pp s s CI J 4Ap S L&4 se*(,'s s.4 as, t A 1-.Ssr 2.3 DATA DZSPIAY SYSTEM I s s I 2.3.1 Technical rt Center Rxxn Each D.C.Cbok Unit has a dedicated corrrrrand console located in'the Onsite Technical Support Center.Each cormend console is ecgupped with two color CRC displays and a video hard copier (which can be used to obtain a hard copy of tt~screen image).One CRT is dedicated to the PSSD function and the second CRP is a general purpose display.Three satellite stations, each with a color CRP display, are also provided.'Ihe satellite stations can be connected to either Ccok Unit 1 or Unit 2 TSC Ccaguter System.A shared video hard ccpier is provided for the three satellite Cps.The satellite stations are arranged so that visual access from the ccrmrand station can be h maintained while still providing sufficient xccm to minimize noise and distrutanoe.

For printing lengttF reports, a line printer is provided.2.3.2 Control Rnn.Two redundant PSSD display CRTs and two redundant BISZ Cps are provided in ea&control room.A video hard copier is also provided to cbtain had ccpy output frcm the CRT screen image.hs s 2.3.3 EE Ooeratin Facilities (EOP): A color CRT terminal, which can be connected to either Ccok unit TSC ccmputer, is provided in the Emergency Qgemtirg Facilities.

'Qm remote CRT can be used to display all of the displays available on AEP-5

~~the PSSD, TSC and BZSZ functions except for the top level iconic display af the PSSD functian.This iconic display was designed for early xeaxpu.tion af an event by tie contxol nxxn cpamtors and therefoxe is not included in N~EOF.

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~Ie~gag gef>>e~y Ct AS I IS5SIOI 5 Ilet!.I.I, IIISLSII I II@~AOeeeeoees I 00 e A'I OII Aef~Ae<<OS Ae ccottelt S~eeet l A CA<<II I~F j I ZOP CH (shared by l+l I I t~I I I I~OSO SI4AACC I I I I I I~I I I~et SAt AIIQOO OlltLA~Ctll I OettlAr<<e<<O l04elc<<(CS Er'v0 Of PlAe ctle I 5 IOOISC IIO ttA<<SIOO~ye000r 0l etc Ar CI<<I AS S At AASIOO I I I I I I I~I I I t I I 3S2$2 00 HDL', 5 I I I 5 l I I I I I 5 COIISOOL LA 5 I I r'I I e A I I I I I I t I I I I I'0&%<<h&W<<CCr SISI Stf SOIAO CAI I III~e'SOAAO teOI af tete~eeeeA OOSA f aos aAA ee eeOAAO Cef Jr40~I eeQAAO'IIOI 0 Ccthl 1 teSO swlAr wlAr CIA~CCR~oveAr CIAO COI I 0 I h<<heO 4 Ah%A I OII OI~IAI IO'%>>4tel A~I<<II<<el~~e0lh<<e~f e<<ehe0~I heOhe0\Set tt I 5 55CIIIIIL'OL Iltfretfl I CIIIII 0~t I l I~I f I I I I I I I I I I t Figurc 2.1.Tt.'chnical SUpport Complex SYstcm Configuration

<<Sensor Signals non-'safety Sa ety syst.syst., sianals sianals iso lato rs<<I<<.~~Control Board Indication I I I I GISX Displays IPSSD Displays I CONTROL ROO14 r L'PSSD Displays Plant Process Computer Isolator X/0 Canine rain A I I I l I L tors~-</O Ca@inc.l Train B I I I I I t e I IBISI Displays I r I I Tsc Displays I TECsi SUPPOFT CENTER TSC CO><PUTTER SYS E.'l site 8 oundary E'igu e 2.2: TSC Computer System Schematic.

TSC BXSI PSSD (non-iconic)NUCLEAR DATA LID K AEP-8 3 ONSITE TE'CHNICAL SUPPORT CENTER 3.1 DESIGN BASIS: Tt~Qnsite Technical Support Center (OTSC)saves as the focal point for post-accident recovery manageaant.

As such, it must have the chili~to access, display and transmit pertinent plant status information independent of actions in the contxol zccm.Technica1 Support Center Su~ion of the TSC Canpxter Systan was j 1.Pexmnnel in the OTSC mat have access to the real time information definix~the jument status of critica1 plant systans and functions.

2.Tfu TSC fur~i mmt have the capability to store historical

~vent and post-event data in order to enab1e a diagncsis and evaluation af th event to deteanine t!m extent af any possible plant system dana<a.3.The TSC Qzmticn nust have the capability to acorns and display plant gararetezs irdependent of actions in th contxol room.4.The interface of tbe TSC system equipment with exisiting plant protection system, control roan or~~func~5.Pazanatars to t?m extent possible should be fran<w sana~e that is used for control rocm irZications to ensure data cons~cy~6.Tlm TSC systan nust have the capability of interfacing with camrnnication equizztant for the offmite tzansaussicn of pertinent.

plant data.

'.'Ihe users mast be able to cr~te or modify displays to naet;tom needs as conditions may dictate.In order to define the information which nust: he available in the OTSC, a generic study af critical plant systems and key safety 8uwtions (as Listed in Table 3.1)was conducted by Westinghouse.

This study resulted in a.List af pazanaters to be monitored by the carputer for the Technical Support Center Suction.This West~ouse paraaater list was reviewed and made~Plant specific by AEP.Table 3.2 Lists the pmnaipal paranatms and Table 3.3 lists the basis for input selection.

Bedtm3ancy and diversity af process ir~tions are utilized to satisfy concerns associated with unavailable signals due to sensor failure.Sana.refixmaent af th input paranaters List may he made after the suhnittaL of'his conceptual design report AEP-10 MESHN6HOUSK NOPRHYARY CIJ55 2 3.3 OTSC OPERATOR INTERFACE The ability of the OTSC to be an effective Mo]fn post-accfdent recovery management is a function of the inputs provided and the abf1fty to present information in a meaningful and organized manner.As stated previously, the man-machine interface fs through the use of fnteractfv'e

~aphic color CRT displays.The interface Anctions fn the OTSC consist of displays and console functions..The display types available for OTSC personnel use consist of graphi'c and alphanumeric displays which are both preformatted and user construc-tible.Examples of the types of dfsp1ays avaf1able are shown fn Figures 3 li 3 2 and 3-3 Figure 3.l.fs an examp]e of a preformatted system status display, g~thering important system and loop parameters onto a sfngle page of display.Figure 3.2 shows more detailed information on individual parameters such as information on sensor status, current~value, and high and low limits..Figure 3.3 is an example of a graphic trend display showing a time history of re]ated parameters.

Highlight--ing techniques for ind~cating parameters vr conditions of)nterest util-.4ze both color and achraaatfc means.By providing a combination of both preformatted and user constructible displays the OTSC personnel are provided with prearranged quickly acces-sfble sys em information and the flexibility to permit the tailoring of information prmentation to meet specific needs as conditions.

dictate.The specific content of preformatted displays will be determined by malyzing pos accident data requirements in terms of event evaluation, the safety situs of the plant, and long-term recovery planning.Ois-plays will also.be designed to ref lee.plant specific design details.8~Oisplay access is provided both by dedicated functional console push-buttons and standard keyboard entries.Ocdicated keys provide access to the most frequently used displays or functions.

For other functions access can be either direct by entering short codes or by utilizing~n instruction func.ion to determine the identification code for a display if it is unknown.~51A 2 7 0

1 gESTI~IGHOUSE PROPRIETARY CLASS 2~Other types of information is available through the console keyboard..These consist of functions such as point revie~, logs, post-trip histor-1cal data review, and offsite data transmission.

The paint review functions enable the console operator,to

'review plant sensor information.

The types of review functions available are: Values of individual points.~2.Points removed from scan.3.Points removed fran limit checking.4 Points failed under quality checking routines.$.Points whose'can frequencies have been changed fmn the normal scan frequencies.

There are log<unctions available to the OTSC personnel which can be displayed on CRTs with periodic updates or output onto a hard copy device such as a line printer.These functions can be preprograrmed and automatically initiated or specified and initiated hy console operator input.I~The post-trip review function provides the capability to review histor-y ical data to aid in an event evaluation.

This function continuously stores in memory an updated table of preassigned sensor values for a, predefined per iod.Upon the occurrence of a disturbance

{e.g., plant trip)the system continues to store data for a defined time period.After this period, the entire, data record can be reviewed by the OTSC personnel on CRTs and/or output to hard copy devices for permanent record storage purposes.CAC1 1'=8 AEP-12 Ip-.,iNGHOJSE PROPRIETARY CLASS Z , The offsite data transmission function enables QTSC personnel to'trans-.mit plant data to offsite',ocations via owner supplied comnunications systems.The OTSC operator can initiate transmission of data either on a"one-shot" or periodic"asis.The transmitted data can be arranged hnto four edited versions for the specific needs of separate offsite:.ccnmunications receivers such as the NRC.\2-9 iI II"5 4 5YZGHGHOUSE PROPRlEFARY CLASS 2 TABLE 3.1 ,4-CRITICAL P'LANT SYSTEMS/FUNCTIONS.Reactivity Control ,Vrimary System Inventory 5 55 5 rI 5 hg c;3 5 5.5 5 Core Heat Removal Capabilities Availability and Capacity of Heat Sinks 5~, Containment Integrity-'Primary System Pressure and Temperature Availability and Capacity of Alternate Rater Sources Availability and Operability of Critical Support Systems-Radioactivity Control 4 5 4~I'5 h'5 5*-2-10 AEP 14

, Table 3.2 TSC Paxarratmrs List Variables Min.No af Signals bet leg tarp-RCS cold leg tarp-BCS pressuxe-Beactor water Level-KS lxxcn concentxation-Pressurizer water Level-Steam generator Level Wide xange Nanna'ancp-Steam Line pressure~ntainFient pxess suxe 0-700 deg F 0-700 deg F 0-3000 psig 0-100%0-5000 pcm 0-100 0 0-100 0 0-100 0 O-a4OO psig-5-+36 psig~denote storacp tank level 2-2oxic acid tank level-Aux feed warm f lac'eed water flow~gh~injection f1cw 589'-599'lev.599'-614'lev.0-100 0 0-100 0 o-aoo~0-250 KLbs/hr 0-5000 K1bs/hr 0-200 cpm AEP-15

~"4 Table 3.2 TSC Paranaters List Vaziab les Min.No of Signals 44 4'4 e-Low head injection flew 4 16~Neutron flux-Contml rod position 53-Prirrary system relief&.4 anent cooling water flow 2~agormnt ccoling water temp.2-Contaimnent targerature 8 0-5500 gptn 0-2500 deg F 0-10000 gptn 32-200 deg F 0-30%0-100 deg F 0-120%pram Pall in or rot Closed-not closed ,4-Sec.syst.relief valves 4-P2R relief~pressure 1-PZR relief~3c 1m'.1-BCS degre of subcooling N/A-Accunulator level-Accunaxlator pressure-AcaxaQator isolation valves 4-Aux building sump level-BHR system flow Closed-not closed Closed-not closed 0-100 psig O-10O S 50-350 deg F 200 sub-5 super 0'-100 8 0-700 psig Closed-r~clcsed 0-f lccd level 0-7000 apn Table 3.2 TSC Paxarretars List Variables Min.No of Sicnmls~heat, ex.outlet temp.~ric acid chaupir@flaw-KS let-dawn flaw-BCS nake-up flaw~xg vBDtilatich dcntKer-Status af standby~-Kigh radioactivity liquid tank level-Badiaactive gas decay tk press 4-Beactor Coolant Punps status 4-PZR neater bank status<<Wtmrolcxy Mind dizection Atm.delta temp.-Badiation 2 Car~ant area xadia~1 1 Containmzt air auriculate QCit Vent radio gas Chit Vent iodine 0-400 deg F 0-10 pe 0-200 gpn 0-200 cd closed-nat closed Emxgized or not 0-100 8 0-150 psig 0-1200 anps 0-200 anps 0-360 deg 0-100 miles/hr 0-50 Peg F.1-10E4 mR/hr 10-10E6~~10-10E6 axn 1O-1OE6~10-10E6 cd AEP-17 Table 3.2 TSC Pazaneters List Variables Min.No.of Si ls-Radiation (continued)

Steam gen.blow down Condenser air ejector Cooling water East Ccoling water West Service water East Service water West Waste Ziquid off-gas Waste gas decay Control rccm area Spent fuel area ClarLzg pp room area 10-10E6 cpn.1-10E4 mR/hr.10-10E6 cpn 10-10E6 cpn 10-10E6 cpn 10-10E6 cpn 10-10E6 cpn 10-10E6 cpn.1-10E4 mR/hr.1-10E4 mR/hr.1-10E4 mR/hr Ncrta 1: Degree of subcooling will he independently calculated by the TSC ccnauter.Note 2: We radiation signals listed above are signals from the'I existirg radiation detectors.

AEP is in the process of irmlementing a new Radiation Ronitor System at Cook Units 1 and 2, and will provide a separate Radiation Data Display System for the TSC and EOF.AEP-18 h Jrh'A~Ah w~hh~r~A 8 hhr 4~PARAHETER tYES1'INAllOUSE PltOPAlEfAQ'LASS 2 TABLE 2 3 I'SC INSTRUHENT BASIS INITIAL EVENT DIAGNOSIS+

u BASIS.(b,c)Containment Pressure Steaml ine Press.ure-Determine if break is inside or outside of contairunent

-Determine if hiqh energy secondary l)ne rupture occurred-Honitor containment conditions

-Haintain an adequate reactor heat sink-Honitor secondary side pressure to:.-verify operation of pressure control steam dump system-monitor RCS cooldown rate Narrow Range Steam Generator Water Level Wide Range Steam Generator Water Level Boric Acid Tank Level Condensate Storage Tank Level Refueling Water Storage Tank Leve I-Determine if malfunction of secondary side system has occurred-None-None-None-None-Honitor heat sink-Haintain steam generator water level-Determine if heat sink is being maintained

-Verify RCS boration system functions for adequate reactivity control-Haintain adequate water supply for auxiliary feedwater pumps-Verify adequate supply of emergency core cooling water-Verify ECCS and containment spray system are functioning

>Certain.indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery, 525lh YIF~SIlNMl""'E I'"ONIFJARY'LASS 2 h

PARAHETER Wide Range Th and Tc-None'PfYiTIHGNONE PAOPAIHAN VMS 2 IAOLE 2-3 (Continued3 TSC INSTRUHENT OASIS INITIAL EVE NT DIAGNOSIS*

I I OASIS (b,c)I-Haintain adequate reactor heat sink-Haintain the proper relationship between RCS pressure and temperature

-verify vessel NDTT criteria-maintain primary inventory subcooled-maintain safe shutdown con-dition-maintain RHR considerations for cooldown-monitor RCS heatup and cooldown rate Pressurizer Hater Level-None-Confirm if plant is in a safe shutdown condition-Determine ability to control RCS pressure-Honitor RCS inventory-Haintain pressurizer water level*Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.5251A Qf'Slla<n<IQI<t

~~no~nI~4n<<

n~"c<<2 0 ICSTIHGIIOUSE I AOPRIDARY CIASS 2 TABLE 2-3 (Continued)

TSC INSTRUMENT OASIS PARAMETER System Wide Range Pressure-None INITIAL EVENT DIAGNOSIS*

BASIS (b,c)-Determine if plant is in a safe , shutdown condition-Maintain the proper relationship between RCS oressure apd tempera-ture-verify vessel NDTT criteria-maintain primary inventory subcooled (particularly with loss of ol'fsite power)-maintain l?NR considerations for cooldown Containment Building Water Level-')etermine whether h'igh energy'line rupture>as occurred inside or outside containment

-Determine NPSll for recirculation mode cooling-Determine which equipment in con-tainment is submerged Condenser Air Ejector Radiation Steam Generator Blowdown Radiation Contaienent Radiation-Determine if steam generator tube leak.has occurred-Determine if steam generator tube leak has occurred-Determine if high energy line break or fuel mishandling accident-Monitor radioactivity release path to environment

-Monitor radioactivity release path to environment

-Moni tor radi oactivi ty release path to environment

-Determine accessibility to con-tainment building*Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incldpnt Recnvery.525lA lVBTlHG!lOUSE

.",".A."."I:.'.r; Cr PSS P MSIIHCIIOIISE PIIOPIIIHAW CLASS 2 TABI.E 2-3 (Continued)

TSC IHSTRIINENT BASIS PARAHETER IHITIAL EVENT DIAGHOSIS*(b,c)-Determine if significant fuel damage has occurred pl&I hJ h)Auxiliary Feedwater Flow Iligh tlead Safety Injection Flow Low llead Safety Injection Flow Area Radiation Honltoring in Auxiliary Building and Control Room-Hone-Hone-Hone-'etermine if source of accident is outside', cont a I runent bu i l ding-Honitor environmental conditions around equipment in containment

-Determine if sufficient flow exists to maintain heat sink-Determine that ECCS is deliyer-ing flow-Honitor ability to keep core covered-Determine that ECCS is deliver-ing flow-Honitor ability to keep core covered-Infer spray operation-Honitor accessibility to plant zones/equipment

-Honitor radioactivity release path to environment

-Honitor effectiveness of cleanup holdup systems-Honitor integrity of.long-term cooling system I*Certain.indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.5251A WESTIHOIIOUSE PIIOPAIETARY CLASS 2 PARAMETER gf.'f tttQIIOUSE I AOPNITARY CLASS g TABLE 2-3 (Continued)

TSB IRSTRBMBRT BASIS INITIAL EVENT DIAG1IOSIS+

BASIS (b,c)[I A I-Honitor habitability of the control room 7 Core Exit Thermocouples Neutron I'lux-None-None-Determine if core is being cooled-Monitor ability of reactivity control systems to keep the core subcritical

-Determine if plant is in a safe shutdown condition I I Degree of SuLcooling of Primary Coolant Primary System Safety and Relief Valve Position Pressurizer Relief Tank Pressure, Temperature, and Level Containment Isolatton Valve Position-None-None-None-None-Haintain adequate reactor heat sink-Haintain safe shutdown condi-tions I-Haintain primary system inventory-Monitor radioactivity release paths into the containment

-Monitor capacity to relieve primary coolant pressure-Monitor radioactivity release paths into the containment

-Monitor radioactivity release paths to the environment

-Monitor status of containment isolation I*Certain tndtcattcns on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.5251A 5f SIINGIIOIISE Pl'OPIllETNY ClASS 2 PAR AHETER Secondary Safety, Reliefs, and Atmospheric Dump Valves-Hone 0 WESTINGIIOUSE PROPRIETARY CUSS 2 TA""E~.3 (Continued)

TSC INSTRUHENT BASIS INITIAL EVENT DIAGNOSIS*(b,c))-Honitor secondary system integrity BASIS-Honitor.radioactivity release paths to the environment Accumulator Tank Level Accumulator Isolation Valve Position RllR System Fits RllR Ileat Exchanger Outlet Temperature Component Cooling Mater Flw and Temperature

-None-None-None-None-None-Honitor primary system inventory-Determine whether the accumulator tanks have in5ected into the RCS-Determine system operation-Hopitor primary system inventory-Honitor core heat removal capabilities

-Honitor core heat removal capabilities

-Honitor system operation of a critical support system*Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.551A SESTINGIIOUSE PROMJETAB'LASS 2

PARAHfTER WESIINGIIOIISE PROPRIETARY CQSS g TABLE 3 3 (Continued3 TSC IHSTRUHfHT OASIS INITIAL f Vf tIT DIAGtlOSIS*

BASIS (b,c)Boric Acid Charging Flow Letdown flow Water Level in Closed Spaces Around Safety fquipmcnt)n Auxiliary Building Emergency Ventilation Damper Position ltigh Level Radioactive Liquid , Tank Level-tlone-None-ttone-Hone-tlone-Honitor pr imary system inventory-Determine boron concentration for reactivity control-Honitor ability to control RCS pressure or pri~~ry system inve,,or-Honitor ability to contr ol RCS pressure-Honitor core heat removal capab il i ty-Determine boron concentration for reactivity control-Honitor environmental conditions around required safety equipment outside of containment

-Ensure proper ventilation to vital areas under post-accident conditions

-Honitor capacity to contain and store radioactive liquids'Certain indications on this table Recovery.5251A I are used as secondary diagnoses as the operator proceeds through Post-Incident gf g]NIIOIISE P.".".0" IQ'ARY Clh.S 2~~

PARAMETER Radioactive Gas lloldup Tank Pressure'tatus of All Electric Power Supplies and Systems Effluent Radioactivity Noble Gases, Radiohalogens, and Particulates Plant and Environs Radioactivity (Permanent and Portable Instruments)

Sampling System Meteorology (wind speed and direction temperature prof lie, and precipitation)

-tlone-None-None-tlone-None Vlf."TltlGllOUSE PROPAIETAAY CLASS 2 TABLE 2-3 (Continued)

TSC INSTRUMENT BASIS INITIAL EVENT DIAGNOSIS*

BASIS (b,c)-Honitor capacity to contain and stare radiaactive gases-Ensure adequate electric power ta safety and suppart systems-Honitor radioactivity release paths to the enviranment

-Monitor release of radioactive materials not covered by effluent monitors-Oetermine RCS chemistry for reactivity control and extent of fuel clad damage-Monitor radioactive effluent transportation for emergency planning, dose assessments, and source estimates Containment Atmosphere temperature

-None and ttydrogen Concentration

-Monitor containment integrity-Honitor environmental conditians around equipment in containment

• Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.5251A VIESTlrlcttOUSE PROP;;tETNW CLASS 2

iNgiiNGHOUSi PROPRIEMRY CLASS 2 Systems Status-Reactor Coolant System Loop 1 Loop 2 Loop 3 Loop 4 T average ('F)Overpower DT PoPWR)Overtemp.DT (%PWR)Cold leg temp.(narrow range)('F)Hot leg temp.(narrow range))'F)Reactor coolant flow (%)Reac'.or coolant pressure-WR (PSlG)Pressurizer pressure (PSlA)Pressurizer vapor temp.(')Pressurizer liquid temp.('F)Pressurizer relief tank pr.ssure (PSlG)Pressurizer relief tank level ('h)Pressurizer relief tank temp.('F)Pressurizer safety relief temp.('F)595.2 595,2 110.0 110.0 110.0 110.0 559.8 559.8 624.0 624.0 100 0'G~0 2250.0 2250.0 2250.0 563.8 565.2 1.5 77.6 110.3 120.0 595.2 110,0 110.0 559.8 624.0 100.0 2250.0 595.2 110.0 110.0 559.8 624.0 1GO.O 2250.0 Figure 3.1'System Status Display at Qnsite Technical Support Center (Example)AZP-27 yIggHGHOUSE PROPRIETARY CUSS Z Parameter Summary Point Qescription Yaiue..Range Units Status TO400 RCS Loop 1 Hot Leg T 593.4 0:700.OEGF Normal PO480 RCS Pressure 2234.1 OOOO LO421 Stm Gen 2 Narrow Range Level 39.1 0:100 PO549 Steamline Pressure LO103 RWSi Level LO114 Boric Acid Tank Level 893.0 0:1100 100.0 0;100 98.8 0:100 LO119 Condensate Storage Tank Level 58.4 0:100 LO947 Containment Bldg.'Vater Level 3.3 0:160.TO406 RCS Loop 1 Cold Leg T 5472 0:700 OEGF.Normal PSlG Normal PC Low PSlG Normal PC Normal PC Normal PC Normal PC High Figure 3.2: Parameter!n'ormation Oisplay at Onsite Technical Support Center (Example)AEP-28 16108-2 WEST)HGHQUSE

?ROPRlETARY CLASS 2 RCS COLD LEG TElNP (oF)100 700 RCS HOT LEG TEMP (4R'100 100 PRZR LEYEL (~o)40 2500 PRZR PRESSURE (PSlG)1900 0 2 4 6 8 10 12 14 16 18 20 T)ME (SECONDS)Rgure 3.3Graphic Display at Onsite Technical Support Canter (Example)AEP-29

/<<'.s'wxA'aa<<~P 4ai r w<<<<<<.,'/<<./

<<-.ms<<w~-/t:.'/ga.~<<aa sm/~as,a/~

wt'<<<<4iv~wm/weaww'<<c4~

V/Ella(GHOIJSK PROFRIEfARY CLASS 2'.0 PLAI'lT SAFETY STATUS DISPLAY 4.S PURPOSE The functian of the Plant Safety Status Display (PSSD)is to present a succinct account of the overall plant safety status to the control room operator (or supervisor).

The entire data base should be available to the operator arranged in a format that will enhance his response to events and the diagnoses of the cause of the event.Because the PSSD serves as an i~a ortant interface between the plant process and the operator, the information presentation should be defined in terms of parameters and logic supportive of defined operating.

procedures for dealing with abnormal events.4.2 INPUT DETERMINATION In urdar tu determine the".squired cperatinna1 mades fnr the PSSD gene must first consider'he various types of transients which may occur.A review af postulated plant transients (events)indicated that they can be divided into two basic categories: (b,c,e)1.Slaw transienats wnich da not result in imnediate protection systems actuation and for which the control room operator has an opportunity to react to possibly terminate the event before safety systems are required to function.Z.Fast transients which result in almost immediate reactor trip and poss'.bly safeguards ac uation and for which the control r oom operator's resporsse is to react to ensure that appropriate safety measures have been taken and to diagnose the event(.Because cf the fact that Ldi-;-erect parameters and signal ranges ara associated with the two potential event typegs the PSSO incorpar ates Ltwo cperating mades.The',ir't made (TERMIRATE MODE)is itive whi1e:hge (b,c,e)!b,c,e)4-1 5435A AEP-30 I IIESTll'lGHGUSE PRQPRlEl'nRY CL(SS 2 At (b,c.e)LpIant is 1n a normal operating cond1tion and the second mode (MITIGATE MDOE)is active following a reactor trgp , The parameters available for'3 (b.c-)Leach mode were chosen to maximize the useful amount of 1nformat1on to be (btc.e)displayed to the operate The role for which the:pSSD providesLsupport for each of the operating mode/a 1s as foll'ows: t (b c e)l cEMIMATE MODE 1.Monitor the plant process for abnormalities indicative of slow transients that do not result in imediate reactor trips and for which the control room operator might take corrective or protective action.2.Monitor the integrity-of the various boundaries to radioactive release.MITI f)ATE KOOE 1.Monitor the safety statu" of the as tripped condition.

2.Monitor for conditions which might lead to a breach of any of the levels of defense against radioactive release.3.Monitor the condition of the barriers to radioactive release.For any event, the safety status of the plant can be evaluated in terms of six basic safety conc ms.These concerns can be stated as follows;1.Saturation of Reactor Coolant 2.Reactivi y Excursion 3.Loss of Primary Coolant Inventory a.Loss of Pressure and Temperature Contre/1 4 2 AEP-31 LB.Radioactive Release (VESTNGHQUSE PRQPRlETARy CLASS 2 (b,c,e)5.Containment Environmengt By addressing Lacy safety concerns, the consequences of abnormal events can be limited or mitigatgd, (b,c,e)tThe tey safety concerns can be related to specific abnormal occur-rences..Tables 4-I and 4-2 indicate key safety goals for some typical postulated events in terms of the PSSD operating mode.It must be noted that these events are typical and it is conceivable for multiple events to occur in undefinable sequences.

For these reasons, the PSSD must be designed on the basis of key safety concerns rather than specific scenario/a

/In defining the inputs for the PSSQ, ttuo requirements have to be me+~tias fo 1 1 ows:.(b,c,e)-(b,c,e)Ll.The inputs selected must represent a minimum sat sufficient for monitoring all possible events including those which might not have been anticipated.(b,c,e)2.The selection of inputs must address conditions with potentially erroneous signals, conflicting indications, and parameters out of range (I.e., redundancy and diversity)g In response to the Lfirst requirement, the function of the PSSO has been considered in two ways.The primary function is to monitor the plant proc ss in terms of satisfying the key safety concerns.As stated above, by guaranteeing that these concerns are addressed, the conditions of unanticipated events or event sequenc s can be satisfied.

The second function of the PSSD is to support the monitoring function of the plant for postulated events and to provide a man-machine interface design that supports a.defined evaluation process and procedures for responding to abnormal events (b,c,e)"43"8 4-3 AEP-32

'NEST!!1B!

HOUSE PROPRIETARY CLA$$2 fn order to satisfy the Lsecond consideration of evaluating erroneous signals and the need for redundancy and"diversity, the PSSO must perform-operations upon multi-sensor inputs to evaluate erroneous signals and be able to provide the operator with a diverse method of indicating the plant process.The inputs to the PSSO are chosen upon the basis of their direct relevance to the key safety concerns.Tables 4-3 and 4-4 list some specific inputs related to key safety concerns for several events 4.3 MAN-i%CHINE INTERFACE (a,b,c)(a,c)The PSSO system will-process the defined input data set of plant param<<, stere atftwo second interval/sand generate displays for redundant PSSD dedicated CRTs located in the control room.QA dedicated CRT will also be located in the Onsite Technical Support Centaur In order to achieve an effective man-machine interface, the display system must be designed to provide a logical and human engineered dis-play structure and selection process in a manner which supports defined roles in which the operator is expected to perform during an abnormal occurrence.(b,c)The role of the control room operator inLdatecting and reacting".o an abnormal occurrence is expected to follow the rour basic activitieQs depicted in Figure 4-1.The display system structure should be.defined such that it Lsupports an identifiable goal for each of the general activities shown in the figure The.se goalgs are defined as follows: IActi vity: Detection Goal: The control room operator should be in a state of readiness to make a rapid detection of incipient threats or actual events which may affect plant safety.The response of the operator would be based upon his knowledge of expected plant performance and his skill in controlling the plant process!.

Nay." 4 Jvsaa's'w 4 a Activity: Reacti on V'-"S IflGHGUSE PROPS!EERY CLASS 2 (b,c)Goal: The control room operator must immediately react to the detection of an event.His irst objective is to assure that appropriate safety system responses have been taken and that key safety concerns are being addressed by observing critical plant parameters.

Activity: Diagnosis Goal: Following the control room operator's inmediate reaction it is then necessary to diagnose the cause{s)of the event and determine if any damage to the various barriers to r adioactive release has occurred.The operational mode at this time would be based on the operator's knowledge supported by reference to various abnormal and emergency operating procedures.

0 Activity: Terminate/Mitigate Goal: At the later stages of the event the control room opera-tor will need to implement the rules or strategies that have been identified as a result of the diagnosis activ-ity.The operator's goal is to verify that corrective actions ara satisfying the key safety concern/a The display structure shovtn in Figure 4-s/supports the specified control room operator activities and goals.The displays are structured into three levels of information ranging from general plant systen sumary information with a broad field of at.ention, secondly to a level of information with a narrower field of attention and more definitive information on subsystems and functions, and finally to a level of information containing irdividual sensor values and statuQs (a,c,f)r,3 4 g AEP-3 4

'hil'G~JSC P Q,la TAR(CLASS QLeveI 1 would contain information in the form of a continuous graphic display for each of the two operating modes of the PSSO.Information contained in the display would support the detection activityI A major problem associated with the man-machine interface is the/requirement that the plant operator sample and process a'large number of plant parameters and perform what are termed multi-parameter decision processes.

An advanced concept in graphic CRT display designed to aid the operator,, is employed for Level 1 information in the PSST Figure 4-3 is an illustration of the display.IEach ray in the figure repre-sents the scale for a process parameter.

When the normal operating values for the parameters are plotted on the scales and lines are drawn connecting the points, a geometric pattern is developed.

Positive deviations from the normal values result in points further away from the cente~of the figure while negative deviations result in points closer to the center of the figure.When the actual values of parameters are different from the normal or reference values, the result is a geometric pattern different from the original patter/a Figures 4-4 and 4-5 are preliminary versions of[Level 1 displays for each of the PSSD operationaI mode/a for two sample events: Primary to Secondary Coolant System Leak and Primary Coolant Systan Leak to Containment.

The parameters chosen for the displays were chosen to/permit an evaluation of the tey safety concern/a/This advanced graphic dispIay provides two distinct advantaoes over conventional control rocm indicators:

a concise, systems level oriented, integration of parameters and secondly, a graphic display format.The detection of an abnormal condition is enhanced as the oper ator task is now based upon the discrimination of two geometric figures.NuIti-parameter decisions and event evaluation is facili-tated by the integrated nature of the display and the fact that only differences in parameters are highlighted by the display.The operator upon detecting abnormalities is then able to se k more specific informa-tion at other information levels to support the reaction, diagnosis, and terminate/mitigate activitiegs 5435A 4-6 AEP-35

)';-ST!,'su,",OUSE PROPrltTAnY CLASS 2 The'inforaation atfLeve1 2 is an expansion of each of the key safety concerns and systems.blare detailed information is provided on the status of the process.For example, the'values of pressures and water levels in individual steam generators could be provided at this level.In addition, trend displays for the previous 5 minutes of operation of Level 1 primary display parameters are prov',ded.

Diversity in process indications at this level will be employed to enable the operator to verify conclusions.

At Level 3, the data is detailed further to provide information on the status of individual sensors, multiple measurement points, and data anomalies.

The sensor values are annotated to include'such things as data-out-of-range and process limits.Information on suspect data qua11ty is carried into upper disp1ay leve1@(a,c,f)543""A 4~7 AZP-3 6

• e e'e jÃESTlHQHOUSE P!OPRfET'qY Ct~SS Z TASLE 4-1 (b,c,e)PLANT SAFETY STATUS..OISPLAY>>

SAFETY GOALS-TERMINATE MOOE TRANSIENTS Reactor Control Systems Malfunction ee e4 e.'*C'~1 C--+i A Stop rod motion Maintain core thermal and nuclear parameters within limits Reactor Coolant System Makeup Control Prevent core thermal and nuclear parameters from exceeding limits e'Maintain-pressurizer pressure and level Inadvertent Oepressurization (Slow)Terminate depressurization Restore systan pressure Reactor Coolant System Leak Limit radioactive release eg Maintain pressurizer pressure and level'e 0 54351 4-8 AEP-37 V/EST1HGHOUSE PRCPRluARY CLASS 2 TABLE 4-2 Reactor Trip PLANT SAFETY STATUS OISPLAY-SAFETY GOALS-MITIGATE MOOE TRANSIENTS (b,c,e)Maintain heat sink via steam generators

-,.Maintain subcooling by controlling steam pressure Maintain pressurizer level Station Blackout Provide secondary heat sink Maintain subcooling Maintain pressurizer level Emergency Eor ation Prevent return to criticality Operation with Natural Circulation Provide heat sink Control subcoo 1 ing Maintain pressurizer level Spur ious Safety Injection Oetermine safety injection is not required and terminate action Loss of Reactor Coolant Verify and establish short term core cooling Maintain long term shutdown and cooling 5435A 4-9 AEP-38 V<ESTtfsGHOUSE P OPHIET: RY CLASS 2 TABLE 4-2.(Continued)

PLANT SAFETY STATUS DISPLAY-SAFETY'OALS

-MITIGATE MODE TRANSIENTS', Loss of'econdary Coolant Establish stabilized reactor coolant system and steam generator conditions Minimize energy release Prevent lifting of.pressurizer safety valves Isolate, auxiliary feed to affected steam.generator Borate to maintain reactor shutdown margin Steam Generator Tube Rupture Minimize radioactive material release Establish feedwater to unaffected steam generators and isolate.faulted unit Maintain residual heat removal capability

-*-Maintain RCS'ubcooling Prevent over-flooding of faulty steam generator V V C'43GA 4-10 AEP-3 9 QV C'C IYBOllCHG"SE FROPRtci~7l'LASS 2 TABLE 4-3 PLANT SAFETY STATUS OISPLAY TERMINATE MOOE PARAMETERS (b,c,e)Variable Transient Reactor Control System Malfunction Reactor Coolant Reactor'akeup Coolant Control System Inadvertent System Malfunction Oepressurization Leak avg ref Rod position Oelta T Startup rate Count rate Pzr.pressur 0 Charging flow Pzr.level Comp.cool H20 rad Containment rad Air eject rad.Blowdown r ad.Cont.humidity Cont.temperatures Cont.oressure Prz.discharge piping temps PRT pressure PRT level PRT temps RCP seal tempera-ture RCP seal flow RCP seal level YCT flow X X X X X X X X X X X X X X X X X.X X X X X X X X X jJ imp 4-11.AEP-4 0

Y~wRCHOVSE PRCPRIETAC CLASS 7 PLANT SAPPY STATUS OISPLAY"IlTTGATE WOE PARAvETERS VW able Trans1cnt Reactor Trip Station Blackout Gnergency Bar'ation Operation with'latural Clrcul atlon loss of Ceo lant Accident Loss af Secondary Coolant Steam Generator Tube Ruature.Reactar trip breaker Startup rate Reutran flux Rod pos1tfon Turbine tHp Blackout signal Tavg (thermacoup 1 es)Rad bottom 3nd.Primary pressure Stcam flex Feed flow Pressurf ter level Care thcrmacoup les Cont.radar at<an Afr Qectar rad$at5an Slowdown radiation Cont.pressure Pri.M.R.tanp.Stcam pressure Cant.sump level Cant.temperature-Cant.huahdl ty Charging flaw S.G.level B.A.tank levei Aux.fmd flaw Sa'law RMST level CST level X X X X I~X X X X X X X X X X X X X X X X X X X X X X X X X X X X X" X X X X X X X X a-lZ.)cSBA AEP-4 1 16708-1 ygggtfGHOUSE PROPRtH'ARY CLOS 2 (a,c)CONDITIONS NORMAL?DETECTION IMPLEMENT RESPONSE IDENTIFY RESPONSE YES SAP ETY LIMITS VIOLATED REACTION NO DIAGNOSE PROBLEM DIAGNOSIS TAKE COR R ECTIV'E ACTION TERMINATE/MITIGATE Figure 4-1.Operator Response Madel AEP-42 16643 10 WESTINGHOUSE PROPRIETARY, CLASS 2 Qrepftie, Olsplay T p C'y X LEVEL 1 OISP LAY Loop, TSAT TAVQ TH TC PR.Press STIjf/FO Flow P Steam RCP's Pressure Relief Vlv.Safety Vlv.Spray Heaters PER.TCS Reactor Coolant Inventory Przr Level Cte.Leatown W.R.TH.TC.Core TC's TSAT.Etc., LEVEL 2 OISP'LAYS Sensors, Comparisons of Redurufant IVleasurernent Error Ctteeks LEVEL 3 OISPLAYS j Inputs Figure 4-2.Display Structure of Plant Status Display AEP-43

WESTlHGHOUSE PROPRlETARY CLASS 2 Pressurizer Pressure (Value)psi Primary Tavg (Value)F Startup Rate (Value)Oec/Min Pressurizer Level (Value)4k/////////Containment Humid (Value)o/o Temp (Value)'F Charging Flow (Value)GPM Steam Gen Level (Value)Io Radiation Contmt Blwdn Air Eject Fig"~4-3.S~pie Display-plant Safety St tus Dl~tay AEp-44

NESTINGHOUSE PROPRIETARY CLASS 2-~k 4 Pressurizer Pressure (Value)psi Primary T avg (Value)'F Startup Rate (Value)Dec/Min Pressurizer Level (Value)%I I I I I I I Containment Humid (Value)'6'emp (Value)'F Charging Flow (Value)GPM Steam Gen Level (Value)%Radiation Contmt Btwdn Air Eject Figure 44.Sample Plant Safety Status Display-Terminate Mode-Primary to Secondary Coolant Sys;em Leak (SG Tube l ak)AEP-45 WESTINGHOUSE PROPRIETARY CLASS 2 RCS W.R.Pressure (Value)psi RCS W.R.Temp (Value)oF (Value)'F Tsat Startup Rate (Value)Dec/Min (a,c,f)4~'\Pressurizer Level (Value)%I I I I I I I rr~r\\\I I I I I I I Containment Pressure (Value)psi R.V.Level (Value)%Steam Gen Level (Value)%Radiation Contmt Bid dn Air Eject Figure 4-K Sample Plant Safety Status Display-Mitigate Mode-Primary Coolant System Leak to Containment mx-46

'EVESTlNQHOOSE PROPRIETARY CLASS 2 5.0.BYPASSED ANO INOPERA8LE STATUS INDICATION FOR.PLANT SAFETY SYSTEMS 5.1 PURPOSE The purpose of the Bypassed and Inoperable Status Indication (BISI)system is to provide the control room operator with a continuous systems level indication of a bypassed or inoperable condition for the systems comprising the engineered safety features.The system considers the, actual status of individual components including systems level bypasses and control room operator entered inputs for components removed from service.5~2 INPUT OETERNINATION Bypassed and inoperable status indication is provided for the systems comprising the engineered safety features and their critical supoort systems.These systems are identified in Table S.l.This table also identifies the types of components for which monitoring is required, the approximate number of each type of component, and the type of status.information needed.This list is generic in nature and will be revised to meet individual plant specific designs.Ie the evaluation of system inputs, the components in each systan are.considered in the light of being in a proper state to perform or supoort the operation of a safety function.The systems level bypass functions that must also be considered are listed in Table: 5.2.In addition to automatically monitored inputs, the system also considers the effect of component or sys.em out of service inputs manually entered by the control room oper ator.5.3 MAN-MACHINE INTERFACE The interface between the operator and this system is provided by redun-dant CRT displays and keyboard consoles located in the control room.Personnel located in the Onsite Technical Support Center will also be AEP-47

~Ve'FSTlhGHOUSE PRO?RlH'hI1Y CLASS 2 able to access the same information.

The 6IGl uti1izes a structured display hierarchy for the operator'.'nterface.

The display hierarchy is shown in Figure 3.1.The primary display, an example of which is shown in Figure 3.2p con-tains the following information for each of the systems comprising the engineered safety features: L Sypassed or inope~able statu" indication for each affected subsystem on either a systems level and/or train level basis.Z.identification of whether the condition is due to the inoperable status of a component or auxiliary support such as cooling water, power supply, tc.Other levels of displays such as shown in Figure.3.3 provide supporting information on individual components within each subsystem and support system.Lnn additional display provides a".abulation of all control room operator entered inputs ror inoperable components for which automatic monitoring can not be accommodated or for which monitoring does not currently exist whenever the status of a system becomes inoperable or bypassed, the ontrol roan operator will be alerted by an audible alarm and the primary display will indicate via video highlighting (e.g., flashing, color change, reverse video, etc.)the affected systen and subsys.em.

The operator can then access supporting displays tc determine the cause of the bypassed or inoperable condition.

The ontrol room operator must acknowledge the abnormal condition in order to silence the audible alarm.Reinstatement of normal systen function wi 11 also generate a different audible signal.Two additional capabilities of the SISI are the timing and test func-tionss AEP 48

WESTltsGHQUSE PI'.OPHl~iARY CL".SS g/The timing i'unction enab'les the control rom aperator ta set up a count-down timing function for a system which is bypassed or inoperable.

An audible alarm would be generated at the expiration of the operator specified time limit.-This feat'ure would aid the control room operator in complying with Technical Specification time limits for systems unavailable for service.(a,c The test function enables the control room operator to test the ef ect on systems level status of a change in component, status prior to chang-ing the component's status.In response to the control room operator entered input, simulating the affect of changing a component's or sys-tem's status, the system determines tne resultant effect on system operability and indicates the result to the central racm operator 3~3 AEP-4 9

• I TABLE S.l-~BYPASSED At10 ItsOP RABL STATUS ItsOICATIOt1 COMPOttEHT INPUTS~t$y,'I~'a o"a ('.c)~Sstem Emergency core cooling.Auxiliary feedwater Comoonents Yalves=Pumps~.Process (level, pressure)Valves.Pumps.Process Status Open/Shut Operable High/L'o~, etc.Open/Shut.Operable Nigh/Low, etc.0 Containment spray'ontainment i so 1 ation Valves Pumps Process Valves Open/Shut Oper able=High/Low, etc Open/Shut Auxiliary power system Breakers'enerators Voltages Open/Closed/Out Operab.l e High/Low'Containment ventilation Yalves Motors Open/Shut Operable Containment hydrogen recombiners Valves Motors Open/Shu Operable Component coo1ing Valves Pumps Open/Shut Oper able Service water Yalves Pumps Open/Shut Operable 3~AEP-50~.'~~o~~~~4~

A WESTINGHOUSE PROPRjETARY CL4SS 2 TABLE 5.2 Y rq BYPASSED AND INOPERABLE STATUS, INOICATION-SYSTEM LEVEL BYPASS FUNCTIONS Safety injection Low pressurizer pressure Low steamline pressure Manual reset~s4 4 Steaml inc isolation Steam dump interlock Steam generator blowdown isolation 3-5 AEP-51 5251A

IESTINGHOUSF.

PROPRIETARY CLASS 2 Operator Inputs Systems Level Status cCCS Hl Head Sl Accumulators Etc.Prfrnary Qteplay Safety lntecuon Continent Spray lOtnersl Pump 1 Ready Pump 2 Out Valve t Open Stthsystern Cont ponent Level Qtapksy Containment Spray Suc pot Comoonent Cooling ESP Power Etc.Support Systornl Component Lovel Pisplay Figure 5.1 Display Structure-8ypasseC and inoperable Status indication AEP-52 IESTINGHpUSE pRppRIETARY CLASS 2 8YPASSED AND INOPERABLE STATUS DISPLAY SYSTEMS Emergency Core Cooling-High Head SI Intermediate Head Sl Low Head SI Accumulators Auxiliary Feedwat r Containment Isolation Containment Spray Contaioment Ventilation Safeguards Power Source Operable Operable Operable Operable Operable Operable Inoperable

-Train A Component Operable Operable Figure.5.2Primary Disofay-Bypaued and inoperable Status indication AZP 53 WESTINGHOUSE PROPRlETARY CLAS)g Train A CONTAINMENT SPPAy Train 8 Train C VLY101 Pump A Suet YLV111 NAOH Supply Pump A Open Open Operable VLY102 Pump A Outlet Closed VLY103 Headr A Outlet VLV121 Recirc A Closed Closed Refueling Water Storage Tank VLV201 Pump 8 Suet VLV21 1 NAOH Supply Pump 8 VLV202 Pump 8 Outlet VLV203 Headr 8 Outlet VLV221 Recirc 8 Open Open Operable Open Closed Closed VLV301 Pump C Suet VLV31 1 NAOH Supply Pump C VLV302 Pump C Outlet VLV303 Headr C Outlet VLY321 Recirc C Open Open Operable Open Closed Closed LS1 CO Level LS101 Level LS1 02 Levei LS103 Level Normal Normal Normal Normal NAOH Spray Additive LS200 Level Normal LS201 Level Normal LS202 Level Normal TS200 Temp Normal TS201 Temp Normal TS202 Temp Normal Figure 5.3 Secondary Display-Bypassed and Inoperable St-tus Information AEP-54 6 TSC ZNSTRUiiENTATZON As described in Section 2, most of the input signals t t ie TSC computer are taken from the existing instruments which also provide signals for the Control Room indicators.

This approach will provide consistent data in both the control room, Onsite Technical Support Center and the EOF.The input signals to the TSC computer therefore have the same high quality, accuracy and reliability as the control room signal.Znputs to the TSC computer provide transformer isolation for all analog input signals and all digital input signals are optically isolated.Zn addition, all signals from the Reactor Protection Channels are taken after the existing safety grade isolators.

The interfacing of the TSC Computer to the existing plant instrumentation was designed so as not to result in any degradation of the control room, protection system, controls or other plant functions.

Any degre'dation that isq noted during checkout and integrated systems testing will be corrected.

AEP-55 7..TSC KWER SUPPLY SYSTEPS 7.1 POWER YO THE TSC CCMPUPER SYSTl24: 1 b g~y (UPS).This UPS system will provide the TSC c~ezs arB pexiphexal egal@~<with a high quality, transient fxe power source.7.1.1 THE UPS SYSTEM: Figuxe 7.1 shows a one-line diagram (schematic) for the UPS system.-The system consists of xedundar&battezy chairs,.battezy, static invextexs, and static txansfer switches.~ruxmal conditions, the battezy charger converts AC to CC and supplies it to the imaxter.'Ihe battexy charger also kems the battezy at, fiQl charge.The invexter.converts the CC to AC in order to supply the 7.1.2 COHSHQ~S CF PCNER SUPPLY INTERRUPTIGH:

thexe is a power xeduction (dip or degradation) or loss (failure)of the AC pcwer souxce, the UPS battezy?eccnes t".w pr'unary source of D" to&a umexter, rather than 51m battezy charger which has lost its normal s~of AC power supply.Tt~h F for a pexicd of 30 minutes.This allows a sufficient tine interval in which a diesel.genexator (badmp AC source)can be made available to provide power to the inverter.In the unli3cely event of loss or AEP-56 TSC POWER SUPPLY SYSTEM (CONCEPTUAL OESIGN)I EMERGENCY SOURCE NORMAL SOURCE BACK-UP'OURCE I NOEP ENOENT 600 VOLT BUS INOEPENOENT 600 VOLT BUS INOEP ENOENT 600 VOLT BUS O'C.C.BREAKER 225A M.C.C.BREAKER 225A M.C,C.BREAKER 225A AUTOMATIC TRANSFER SWITCH 260A 600'5KVA 120~700 A~MP I BATTERY CHARGER I (ALTERNATE}

700 AMP BATTERY CHARGER BATTERY 927A 40KVA INVERTER 40KVA INVERTER STATIC SNITCH STATIC SWITCH FIGURE 7.I UNIT W I TSC COMPUTER 8 P'ERIPHERALS UNIT W2 TSC COMPUTER 8 P ERIP HERA LS AEP-56o

unavailability af both the rurmal and badcup AC sources, the static swi~will beused for transfer, if necessary, to the enaxcpncy AC source~7;2 PONER TO THE TSC CDMPLZX: Standard balan-plant (BOP)sources will provide the TSC with power for lighting and cowmnience receptacles.

For additional protection,, the lightizq fixtures are provided with battery pactum for continued operation in the event of loss af the EOP pamr supply.The PRC equitant will be supplied frcm an Essential Services System bus QC source).AEP-57 Section 8.0 Original pages AEP-58 through AEP-62 have been deleted from this submittal.

The descriptive information that was contained therein can be found in the OCCNP Emergency Plan.L?af IC 5~C AEP-58

Section 9.0 Original pages AEP-63 through AEP-65 have been deleted from this submittal.

Listings of plant records, plant specific reference material, general technical reference material, plant procedures and reports that are available to personnel working in the TSC are provided in general company internal documents which pertain to the subject matter.AEP-63 Attachment 1 to AEP:NRC:0916I REASONS AND 10 CFR 50.92 ANALYSES FOR CHANGES TO THE DONALD C.COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATIONS I I I j,r/I I 1 rr+ktr'fl g I P Jf en rr~Ab::.,i.j f",",1;i,'j,-';:;,,;

e'I t 4 I).r., g~A rtl 11 I 4~'1 I C jI Ip I)1., j I I'I 1'I 11 IJ I ,4 i I, I rt'>>jar'lr rjh/tr I Il AEP:NRC:0916I Attachment 1 Page 1 of 18 The Technical Specification (T/S)changes included in this letter are, in general, those necessary to support the safety analyses performed by Exxon Nuclear Company (ENC)for the Unit 2 Cycle 6 reload.In addition to these changes, however, we have included additional changes which are intended to make the T/Ss clearer, easier to use, or more consistent with the Standard Technical Specifications (STSs)for Westinghouse Pressurized Water Reactors, NUREG-0452, Rev.4 (or Draft Rev.5, where applicable).

A summary of the changes has been included as Attachment 10 to this letter.It includes a brief description of each change, as well as the reason for the change, and, where applicable, references to the safety analyses the change is based on.This attachment includes an overview of the changes, as well as our 10 CFR 50.92 justifications for no significant hazards consideration.

Please note that the changes will be referred to by their numbers, which are given in the"Description of Change" column in Attachment 10.We have grouped the changes into 12 separate types for ease of discussion.

These changes are discussed below.1.Editorial Changes The first group of changes to be discussed consists of those that are purely editorial in nature.These changes are numbered 1, 2, 5, 6, 12, 20/21'4'5'6'5'0~

60~62~69'4'1~83~84'8'" 90'3J 94'7/98'nd 105 in Attachment 10.These changes are proposed to enhance the readability of the T/Ss, to achieve consistency between the Unit 1 and 2 T/Ss, or to achieve consistency with the STSs, as described in Attachment 10.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: '(2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 These changes, being editorial in nature and intended to improve the readability of the T/Ss, will not reduce in any way requirements or commitments in the existing T/Ss.Thus, no increase in the probability or consequences of a previously evaluated accident would be expected.Criterion 2 These purely editorial changes will not create the possibility of a new or different kind of accident from any previously evaluated, because all accident analyses and nuclear design bases remain unchanged.

AEP:NRC:0916I Attachment 1 Page 2 of 18 Criterion 3 The proposed amendment will not involve a significant reduction in margin of safety, because, as discussed above, all accident analyses and nuclear design bases remain unchanged.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The first of these examples refers to changes.that are purely administrative in nature: for example, changes to achieve consistency throughout the T/Ss, correction of an error, or a change in nomenclature.

This group of proposed changes is intended to achieve consistency between the Unit 1 and 2 T/Ss, to achieve greater consistency with the STS format, or to improve the overall readability of the T/S document.As these changes are purely editorial and do not impact safety in any way, we believe the Federal Register example cited is applicable and that the changes involve no significant hazards consideration.

2.Removal of 3-Loo Technical S ecifications A second category of changes involves removal of Technical Specification provisions for 3 reactor coolant loop operation in Operational Modes 1 and 2.These are changes numbered 3, 7, 16, 29, 30, 31, 46, 56, 59, 61, 67, 91, 99, and 100 in Attachment 10.This category includes all changes involving removal of 3-loop provisions except for those associated with Functional Unit l.e.(Differential Pressure Between Steam Lines-High) on Engineered Safety Features (ESF)Actuation Instrumentation Table 3.3-3.Three-loop changes associated with this ESF signal are discussed in Category 5 of this Attachment.

License Condition 2.C.3(j)for Unit 2 prohibits operation with less than 4 pumps at power levels above the P-7 permissive (approximately 11%of rated thermal power).As a matter of practice, we have extended this restriction to cover all of Modes 1 and 2.As T/Ss covering 3-loop operation in Modes 1 and 2 are therefore not necessary, we propose to remove them to streamline the document.Included in this group of changes is the deletion of T/S 3/4.4.1.4.

Although this specification contains provisions for less than 4-loop operation in modes other than 1 and 2, the requirements for other modes which remain applicable are addressed identically in other T/Ss, as specified below: Action Statement (Below P-7)Where Addressed a b c T/S 3'.1.1 T/Ss 3.4.1.2 and 3.4.1.3 Not needed, since 3-loop operation in Modes 1 and 2 will be prohibited.

AEP:NRC:0916I Attachment 1 Page 3 of 18 Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 This group of changes will extend the license condition prohibiting 3-loop operation above the P-7 permissive to include all of Modes 1 and 2.Thus, the changes would be expected, as a minimum, to reduce the probability, or consequences of a previously evaluated accident.Criterion 2 Since these changes place additional restrictions on plant operation, they would not be expected to create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 Since 3-loop operation in all of Modes 1 and 2 will be prohibited, additional margin to DNB under accident conditions should result.Thus, margin of safety should be increased rather than decreased.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The second of these examples refers to changes that impose additional limitations, restrictions, or controls not presently included in the T/Ss.Since prohibition of 3-loop operation in Modes 1 and 2 constitutes a restriction which the current T/Ss do not have, we believe this example is applicable and that the changes involve no significant hazards consideration.

3.Additional Restrictions Because of Safety Analyses A third group of changes involves inclusion of proposed new requirements in the T/Ss.The new requirements are proposed to make the T/Ss consistent with the safety analyses performed by ENC in support of the Cycle 6 reload, or to achieve consistency with the STS.These changes are numbered 9, 22, 51, 52, 55, 63, 64, 70, 72, 73, 80, 82, 86, 92, and 102 in Attachment 10.The applicable references to the safety analyses are included there also.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (1)involve a significant increase in the probability or consequences of an accident previously evaluated, AEP:NRC:0916I Attachment 1 Page 4 of 18 (2)create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3)involve a significant reduction in a margin of safety.Criterion 1 These changes constitute additional restrictions on the plant in terms of T/S mode applicability, surveillance requirements, or Action Statement requirements.

Since none of these changes reduce in any way previous safety requirements, they would not be expected to result in an increase in the probability or consequences of an accident previously evaluated.

Criterion 2 These changes will place additional restrictions on plant operation and will increase, rather than reduce, requirements for safety.Therefore, they should not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 These changes add additional safety requirements, and in no way reduce any existing requirements.

Thus, no reduction in margin of safety will occur because of these changes.Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The second of these examples refers to changes that impose additional limitations, restrictions, or controls not presently included in the T/Ss.These changes impose additional restrictions on the plant for consistency with the Cycle 6 safety analyses or the STSs.Thus, we believe that this example is applicable and that the changes involve no significant hazards consideration.

4.Refueling Water Storage Tank Chan es A fourth group of changes involves T/Ss 3.1.1.3, 3.1.2.3, 3.1.2.5, 3.4.1.2, 3.4.1.3, and 3.9.8.1 specifically as they apply to borated water addition or positive reactivity addition from the Refueling Water Storage Tank (RWST).These are changes numbered 25, 26, 27, 87, 89, and 104 in Attachment 10.T/S 3.1.1.3 requires reactor coolant flow of at least 3000 gpm during dilution of the Reactor Coolant System (RCS)boron concentration in any mode.T/Ss 3.4.1.2 and 3.4.1.3 require at least one coolant loop to be in operation during boron dilution in Modes 3, 4, and 5.T/S 3.9.8.1 requires 3000 gpm of coolant flow via the Residual Heat Removal System during boron dilution in Mode 6.T/Ss 3.1.2.3 and 3.1.2.5 prohibit positive reactivity addition in Modes 5 and 6 with charging pumps or boric acid transfer pumps inoperable, respectively.

Because of concerns with literal T/S compliance, questions have arisen as to the applicability of these specifications during the times when we add water to the RCS from an operable RWST, specifically when the boron concentration of the RWST is lower than the RCS.

AEP:NRC:0916I Attachment 1 Page 5 of 18 The RWST minimum boron concentrations stated in the T/Ss were established to ensure that adequate shutdown margin is maintained, and are consistent with numbers assumed by ENC in their Cycle 6 reload analyses.Because of this, it is our belief that the boron dilution restrictions of the T/Ss listed above were not meant to be applicable during water addition from the RWST, provided the boron concentration in the RWST exceeds the minimum requirements stated in the T/Ss.We have documented this interpretation in the past (see our letter AEP:NRC:0975A, dated February 28, 1986);this change is submitted only to formalize this interpretation.

Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated; create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 Our review has determined that the T/S RWST minimum boron concentrations are sufficient to ensure that adequate shutdown margin is maintained throughout the entire core life.Additionally, the RWST boron concentrations are consistent with those assumed in the LOCA analyses performed by ENC.Thus, we conclude that these changes will not significantly increase the probability or consequences of an accident previously evaluated.

Criterion 2 The proposed amendment will not create the possibility of a new or different kind of accident from any previously evaluated.

It has been determined that the RWST boron concentration is sufficient to ensure adequate shutdown margin from all expected operating conditions.

The consequences of adding water from an operable RWST which is at a lower boron concentration than the RCS is therefore bounded, and no new or different kind of accident from those previously evaluated would be expected.Criterion 3 Because these changes lessen operating restrictions, it can be expected that a reduction in safety margin may occur.However, because the RWST minimum boron concentrations are sufficient to provide adequate shutdown margin from all expected operating conditions, this reduction in safety margin would be insignificant.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The sixth of these examples refers to changes which may result in some increase to the probability of occurrence or consequences of a previously analyzed accident, but where the, results are AEP:NRC:09161 Attachment 1 Page 6 of 18 clearly within limits established as acceptable.

As discussed above, these changes relax requirements related to boron dilution or positive reactivity addition, but are clearly bounded by our shutdown margin analyses.Thus, we conclude that the example cited is applicable and that the changes involve no significant hazards considerations.

5.Changes to the Differential Pressure Between Steam Lines-High ESF Actuation Signal The fifth group of proposed changes involve Functional Unit l.e (Differential Pressure Between Steam Lines-High) under the Engineering Safety Feature (ESF)Actuation System Instrumentation Table 3.3-3.These changes are numbered 67, 68, and 71 in Attachment 10.Specifically, we are proposing to change the footnote designator for the Channels to Trip column of the 3-loop section to a quadruple pound sign, and to add a corresponding new footnote to the Table 3.3-3 notations on T/S page 3/4 3-21.Additionally, we propose to revise the functional unit to prohibit 3-loop operation in Modes 1 and 2, consistent with Category 2 of this attachment.

The Differential Pressure Between Steam Lines-High actuation differs from other ESF actuation signals in that a signal from one loop is compared to signals in the other loops.The current footnote associated with this signal for the 3-loop case states: "The channels associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode." This could be construed to mean that all channels in the out of service loop should be tripped.This in turn would result in an ESF actuation.

It is our belief that the footnote as applied to this functional unit means to trip the bistables which indicate low active loop steam pressure relative to the idle loop.This action reduces the ESF actuation logic for the active loop differential pressures from 2 out of 3 to 1 out of 2, and thus permits 3-loop operation in Mode 3 since 2 channels per steam line are necessary for a trip.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.The prohibition of 3-loop operation in Modes 1 and 2 is consistent with the changes included in Category 2 of this attachment.

The 10 CFR 50.92 analysis is thus identical and will not be repeated here.The 10 CFR 50.92 analyses included in this category are therefore only those involved in rewriting the Differential Pressure Between Steam Lines-High footnote in T/S Table 3.3-3.Criterion 1 The changes included in this group are editorial in nature, intended only to clarify the ESF Actuation System Instrumentation Table (3.3-3)as it AEP:NRC:0916I Attachment 1 Page 7 of 18 applies to the Differential Pressure Between Steam Lines-High actuation signal.Thus, no significant increase in the probability or consequences of a previously evaluated accident should occur.Criterion 2 The proposed amendment will not create the possibility of a new or different kind of accident from any previously evaluated because these changes, being editorial in nature, will not impact existing safety analyses or the nuclear design bases.Criterion 3 The proposed amendment will not involve a significant reduction in margin of safety because, as discussed above, all accident.analyses and nuclear design bases remain unchanged as a result of these proposed T/S changes.Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The first of these examples refers to changes that are purely administrative in nature: for example, changes to achieve consistency throughout the T/Ss, correction of an error, or a change in nomenclature.

This group of proposed changes is intended only to clarify the T/Ss, to avoid the possibility that they may be misread.As these changes are editorial and do not impact safety in any way, we believe that the Federal Register example cited is applicable and that the changes involve no significant hazards consideration.

6.Changes to the Power-0 crated Relief Valve (PORV)S ecification, 3/4.11.4 The sixth group of proposed changes involve a redraft of T/S 3/4.11.4, concerning the Pressurizer Power-Operated Relief Valves (PORVs).These changes are number 95 in Attachment 10.Specifically, we are proposing to change T/S 3/4.11.4 to require that at least 2 PORVs be available in Modes 1, 2, and 3.For purposes of this specification,"available" means that the PORV is operable with its solenoid deenergized and that the block valve is operable and energized.

This differs from the present T/S, which allows all 3 PORVs to be inoperable, provided their associated block valves are closed.The proposed changes are intended to ensure that PORV relief capability is available to assist in RCS depressurization following a steam generator tube rupture without offsite power, and to respond to comments made by members of your staff at a meeting held with us in Bethesda, MD on December 13, 1984.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.

AEP:NRC:0916I Attachment 1 Page 8 of 18 Criterion 1 This group of changes constitutes additional restrictions placed on PORV (and associated block valve)operability requirements.

Since no restrictions associated with the PORVs are reduced in any way by this group of changes, we conclude that these changes will not increase the probability or consequences of a previously analyzed accident.Criterion 2 Since these changes place additional restrictions

'on plant operation and in no way reduce present safety restrictions, they would not be expected to create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 These changes add additional restrictions on the PORVs, designed primarily to ensure that PORV relief valve capability is available to assist in RCS depressurization following a steam generator tube rupture.Thus, these changes would be expected to increase, rather than decrease, safety margins.Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The second of these examples refers to changes that impose additional limitations, restrictions, or controls not presently included in the T/Ss.Since this group of changes will require PORVs to be operable in Modes 1 through 3 (where previously no operability requirement existed), they clearly constitute additional restrictions.

Thus, we conclude that the example cited is applicable and that no significant hazards are involved.7.Addition of T/S 4.0.4 Exem tions The seventh group of proposed changes are those which add T/S 4.0.4 exemptions to existing T/Ss.These changes are numbered 44, 65, 66, and 103 in Attachment 10.For the first of these changes, a T/S 4.0.4 exemption has been proposed for the flow measurement performed after each refueling and for all flow surveillances for the DNB T/S, 4.2.5.1 (see numbers 44 in Attachment 10).(The flow specification has been moved from the F specification (3/4.2.3)to the DNB specification (3/4.2.5.1) for consistency with Unit 1 specifications.)

This exemption is required H because flow is measured using secondary calorimetric and primary temperature measurements, which can only be performed at or near full power.The flow instrumentation is calibrated based on this measurement.

Exemptions have also been provided for several Nuclear Instrumentation System (NIS)calibrations (see numbers 65 and 66 in Attachment 10)in T/S Table 4.3-1.Of these, those proposed for source range and intermediate range detector calibrations appear in STS, Rev.4.STS, Rev.4 also provides this exemption for the incore detector, excore power range

AEP:NRC:09161 Attachment 1 Page 9 of 18 detector cross-calibration performed after refueling.

Our proposal extends this exemption to the quarterly incore detector, excore power range detector cross-calibration in order to address the situation where an unscheduled outage of significant duration causes the surveillance interval for this calibration to lapse.This exemption is proposed for the daily power range, neutron flux heat balance because it is required to be performed above 15%rated thermal power by T/S.It is also proposed for the monthly incore-excore axial offset comparison for the same reason.These exemptions are needed to address unscheduled outages for which the surveillance interval has lapsed.An exemption from T/S 4.0.4 for the source range channel functional test is proposed.This exemption addresses the situation that results from a reactor trip after continuous power operation of more than 1.25 times 31 days.This surveillance cannot be performed at power without damaging the source range detectors.

Exemptions from T/S 4.0.4 are proposed for the single-loop and two-loop loss-of-flow trip calibrations of T/S Table 4.3-1.These are required because these calibrations are based on the primary flow measurement taken at or near full power which was discussed above in relation to flow instrumentation.

These changes are numbered 65 and 66 in Attachment 10.Exemptions from T/S 4.0.4 are proposed for the f(D, I)penalties associated with the Overpower 5 T and Overtemperature b,T trips.These exemptions are required because the f(5 I)module is calibrated to data obtained from the incore detector, excore power range detector cross-calibration.

As is implied by the exemption of this calibration from T/S 4.0.4 on a refueling frequency, which is already available in STS, Rev.4, this calibration must be performed at power, in the applicable mode.The calibration is performed at power so that an appreciable signal can be obtained on the incore detectors and the excore detectors.

These changes are numbered 65 and 66 in Attachment 10.Lastly, an exemption from T/S 4.0.4 is proposed for Surveillance 4.7.1.5 (see number 103 in Attachment 10.)This exemption is required because T/S 3.7.1.5, Steam Generator Stop Valves, is applicable to Mode 3, and Surveillance 4.7.1.5, which measures stop valve closure time, must be performed in Mode 3.In order to demonstrate the required closure time for the steam generator stop valves, steam pressure must be in the normal operating range corresponding to primary temperature above the P-12 setpoint.Therefore, secondary pressure for this test must be above approximately 800 psig for which saturation temperature is well above the 350 F Mode 3 boundary.An exemption is also proposed for Beginning of Cycle to enter Mode 2 for physics testing provided the steam generator stop valves are closed.This provision allows continuation of the startup program with steam generators isolated in the event that secondary side work is not complete.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (1)involve a significant increase in the probability or consequences of an accident previously evaluated, AEP:NRC:0916I Attachment 1 Page 10 of 18 (2)create the possibility of new or different kind of accident from any accident previously analyzed or evaluated, or (3)involve a significant reduction in a margin of safety.Criterion 1 The changes in this section are necessary to make the T/Ss accurately reflect limitations associated with surveillances which must be performed in the applicable mode.Additionally, the changes are needed to address the fact that unscheduled outages can and do occur, and when they do surveillances can expire with no way to correct the situation until the unit returns to power.Where possible we have followed the guidance given by the STSs, expanding it as necessary to address the situations just described.

As these changes are consistent with the guidance provided by the STSs, we believe that any increase in the probability of occurrence or consequences of an accident previously analyzed, or any reduction in margins of safety, would be insignificant.

Criterion 2 Since these changes require neither physical changes to the plant nor changes to the safety analyses, it is concluded that they will not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 Please see our discussion on Criterion 1, above.Lastly, we note that the Commission has provided guidance concerning the determination of significant, hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

Example 6 refers to changes which may result in some increase to the probability or consequences of a previously analyzed , accident, but where the results of the change are clearly within acceptable limits.It is our belief that these changes are necessary to reflect limitations inherent in surveillance testing methods employed by the Cook Plant, and the changes reflect further clarification of the intent of the original T/S as is indicated by the type of T/S in these areas that is permitted by later revisions of the STS.In light of this, we believe the reasons for this group of changes to be consistent with Example 6.8.Changes to Existing T/S Values The eighth group of proposed changes involve values of parameters presently included in the T/Ss that are being revised to reflect the assumptions used in the various safety analyses performed in support of the Unit 2 Cycle 6 reload.These changes are numbered 4, 8, 10, 11, 13, 14'5, 17, 18, 19, 23, 28, 34, 40, 42, 47, 48, 49, 54, 76, 78, 79, and 101 in Attachment 10.That attachment also includes references to the specific sections of the accident analyses on which the changes are based.

AEP:NRC:0916I Attachment 1 Page 11 of 18 Two types of changes included in this group need further explanation.

The~first are changes to allowances to permit operation with RdF RTDs.These are included in the changes numbered 8, 10, 14, 19, 42, 47, 48, 76, and 78 in Attachment 10.During the Unit 2 Cycle 6 refueling outage, we will be replacing all of our existing Rosemount RTDs with RTDs manufactured by the RdF Corporation.

Because the uncertainties associated with these new RTDs are different from those associated with the older Rosemount RTDs, it is necessary to revise some T/S values accordingly.

We used the revised uncertainties to obtain Technical Specification setpoints from the analysis values calculated by Exxon Nuclear Company.Certain setpoints were affected by both a change in analysis value and the revised allowances.

For your convenience, we have included the Westinghouse Electric Corporation safety evaluation for the RdF RTD installation (WCAP-11080) as Attachment 3 to this letter.The second group of changes needing clarification are changes involved with the f(5 I)penalty which is applied to the Overtemperature 5 T and Overpower 5 T reactor trip setpoints.(These are changes numbered 15 and 18 in Attachment 10.)There is only one f(~I)module, which serves both of these trips.This module places a penalty on these trip functions in the event of an axial imbalance in neutron flux between the top and bottom halves of the core.The f(~I)penalty was not required as an input to the Overpower L T trip for previous Unit 2 cycles, and thus f (L I)is presently set equal to zero in T/S Table 2.2-1.The new analyses performed by ENC apply the f(5 I)penalty to both Overpower and 2 Overtemperature 5 T.The ENC analyses resulted in different f(5 I)functions for these two trips.However, because they share the same f(~I)module, a single f(5 I)function that conservatively bounds these two functions was chosen for the proposed T/Ss.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (1)involve a significant increase in the probability or consequences of an accident.previously evaluated, (2)create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or'(3)involve a significant reduction in a margin of safety.The changes included in this group are necessary to support safety analyses performed by ENC and Westinghouse Electric Corporation (as referenced by Attachment 10)in support of the Cycle 6 reload.These analyses have not yet been accepted by the Commission.

Our conclusion of no significant hazards considerations, which is supported below, is therefore contingent upon Commission acceptance.

Criterion 1 The safety analyses performed for Cycle 6 addressed all previously analyzed accidents.

The analyses, which are referenced in Attachment 10, demonstrated that no sig'nificant increase in the probability or consequences of a previously evaluated accident is expected to occur.

AEP:NRC:0916I Attachment 1 Page 12 of 18 Criterion 2 The safety analyses performed for Cycle 6 addressed all applicable accidents found in the Standard Review Plan for relevancy to Cook.Many of those addressed had not previously been evaluated for D.C.Cook Unit 2.Therefore, we conclude that, to the best of our knowledge, this group of changes will not create the possibility of a new or different kind of accident from any accident previously analyzed.Criterion 3 The safety analyses performed for Cycle 6 (as referenced by Attachment 10)have demonstrated that acceptable margins of safety are maintained for all accidents which were addressed.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The second of these examples refers to changes resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved.These changes are similar to this example in that the Cycle 6 reload is very similar to previous reloads in terms of enrichment, power distribution, and fuel type.Although minor changes have occurred (e.g., F was increased from 2.04 to 2.10), the changes were analyzed and found n8t to significantly impact applicable margins to safety.Thus, we conclude that the example cited is relevant and that no significant hazards consideration is involved.N 9.Se aration of Flow Rate and F The ninth group of changers involve revisions to T/S 3/4.2.3, Nuclear Enthalpy Hot Channel Factor (F).These changes are numbered 41, 42, 43, 48 in Attachment 10.In the present T/Ss, RCS flow rate and F may be hH"traded off" against one another (i.e., a lower measured RCS flow rate is acceptable provided F is also acceptably lower).In the proposed TgS 3/4.2.3, we have eliminated the ability to trade off flow for F.F is hH now defined in T/S 3.2.3 only as a function of rated thermal power.RS flow rate in Mode 1 has been moved to proposed T/S 3/4.2.5.1, which contains the Mode 1 DNB parameters.

Although the Action Statements and surveillance requirements have been revised to reflect this separation, no requirement appropriate for either of the two has been deleted or made less severe.No flux mapping is requiged in the DNB Action'tatement, because flux mapping is used to measure F<, not flow.The proposed changes included in)his group are only those changes involved in separating flow rate and F in the T/S.Changes to existing fH T/S values for flow are included in Category 8 of this attachment.

Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (1)involve a significant increase in the probability or consequences of an accident previously evaluated, AEP:NRC:0916I Attachment 1 Page 13 of 18 (2)create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3)involve a significant reduction in a margin of safety.Criterion 1 This group of proposed changes in no way removes or reduces any safety requirements, nor does it require physical changes to the plant.Thus, it is not expected to involve a significant increase in the probability or consequences of a previously evaluated accident.Criterion 2 These proposed changes will not create the possibility of a new or different kind of accident, from any previously analyzed, because, being primarily editorial in nature, they impact neither the accident analyses nor the nuclear design bases.Criterion 3 The proposed changes will not involve a significant reduction in margin of safety, because, as discussed above, all accident analyses and nuclear design bases remain unchanged.

Since these changes actually represent additional restgictions (in that we will no longer be able to trade off RCS flow rate for F)it could be anticipated that an increase, rather than decrease, in the margin to DNB under accident conditions might actually AH result.Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The first example refers to purely administrative changes to the T/S: for example, changes to achieve consistency throughout the T/Ss, correction of an error, or a change in nomenclature.

These changes are similar to this example in that RCS flow rate and F are being separated with no reduction in requirements, primarily to make Ke Unit 2 T/Ss more similar to those for Unit 1.The second example published in the Federal Register refers to changes that constitute additional limitations, restriction's, or controls not presently included in the T/Ss: for example, more stringent surveillance requirements.

These changes are similar to this example ig that we will be prohibiting ourselves from trading off RCS flow rate for F<For the reasons provided above, we conclude that the examples cited are xelevant and that this group of proposed changes involves no significant hazards consideration.

10.Chan es to the P-12 Interlock Descri tion The tenth group of proposed changes involves the P-12 Interlock description included in T/S Table 3.3-3.These changes are numbered 75 and 77 in Attachment 10.The P-12 Interlock receives input from the T low-low bistables.

These bistables are calibrated to trip when the ave 0 temperature decreases to 541 F as specified in T/S Table 3.3-4.

AEP:NRC:0916I Attachment 1 Page 14 of 18 With 2 out of 4 bistables tripped, P-12 permits the manual block of the Low Steam Line Pressure Safety Injection, causes steam line isolation under conditions of high steam flow, and removes the arming signal to condenser steam dump.With 3 of 4 T channels above the reset point, which is greater than 541 F, the manual.block of Low Steamline Pressure 0 ave Safe'ty Injection is defeated or prevented and the condenser steam dump is enabled.The present T/S description of the P-12 Interlock is confusing in that it neglects the trip and reset points, and instead describes P-12 in terms of conditions above 544 F and below 540 F.If this description is read 0 0 literally, it could be inferred that P-12 is established when T is 0 o ave greater than or equal to 544 F and when T is less than 540 F.ave Additionally, the manual block of safety xn3ection actuation would not be 0 0 permitted until below 540 F, when in fact the setpoint is 541 F.We propose to rewrite P-12 in terms of the 541 F setpoint, which is similar to 0 the methodology utilized in Rev.4 of the STS, in order to better reflect the functioning of this interlock.

In addition to the changes described above, we have revised the P-12 function description.

The current, description states that the Safety Injection associated with P-12 occurs on high steam line flow and low steam line pressure.The D.C.Cook Unit 2 ESF design provides a Safety Injection on Low Steam Line pressure which does not require a coincident signal from P-12 Low Low T.This particular Safety Injection may be blocked if the P-12 Low Low f.'ignal is present.High steam line flow ave ave coincident with P-12 Low Low T does not provide a Safety Injection; it ave does however cause a steamline xsolation.

Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 These changes, being editorial in nature and intended only to more accurately describe the functioning of the P-12 interlock, will not, reduce in any way requirements or commitments which are presently included in the T/Ss.Thus, no increase in the probability or consequences of a previously evaluated accident would be expected.Criterion 2 These changes, being purely editorial, will not create the possibility of a new or different kind of accident from any previously evaluated because all accident analyses and nuclear design bases remain unchanged.

AEP:NRC:09161 Attachment 1 Page 15 of 18 Criterion 3 The proposed amendment will not involve a significant reduction in margin of safety, because, as discussed above, all accident analyses and nuclear design bases remain unchanged.

Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not, likely to involve significant hazards consideration.

The first of these examples refers to changes which are purely administrative in nature: for example, a change to achieve consistency throughout the T/Ss, correction of an error, or a change in nomenclature.

This group of proposed changes is similar to this example in that the changes are purely editorial, intended to make the T/Ss more accurately reflect the functioning of the P-12 interlock.

No physical changes to the plant or its procedures will be necessary because of these changes.Thus, we conclude that the example cited is applicable and that this group of changes involve no significant hazards consideration.

11.Sim lifications to Power Distribution and APDMS T/S The purpose of the eleventh group of proposed changes is to delete reference to the Axial Power Distribution Monitoring System (APDMS)from the T/Ss and to simplify the Power Distribution Limits T/Ss.These changes are numbered 32, 33, 37, 38, 39, 53, and 85 in Attachment 10.The APDMS is an option currently provided in the T/Ss.It is required to be operable by T/S 3.3.3.7 when it is being used for monitoring axial power distribution.

Power operation is permitted above the Allowable Power Level (APL)and below Rated Thermal Power provided additional surveillance is performed using the APDMS in accordance with T/S 4.2.6.1.In practice, however, the APDMS can be somewhat more limiting than APL.More importantly, experience has shown that APDMS causes extensive wear and tear on the Movable Incore Detector System, which the APDMS uses for data acquisition.

This effect results in serious maintenance problems on a system which contains parts which are highly radioactive.

For these'easons, it was decided not to operate with APDMS.Therefore, we are proposing to delete T/S 3/4.3.3.7, and to revise T/Ss 3/4.2.2 (F (Z))and 3/4.2.6 (Axial Power Distribution) to remove material related to APDMS.In conjunction with the above, we have rewritten T/S 3/4.2.6.The proposed T/S contains the limits and surveillances required to establish and maintain APL, and has also been renamed accordingly.

Most of the surveillance requirements of T/S 4.2.2 have been moved to T/S 4.2.6 in order to further simplify these T/Ss.It should be noted that the 2%penalty applied to F (Z)for increasing F by T/S 4.2.2.2.e has been incorporated into the definition of APL in%he proposed T/S 3.2.6.No Q requirements or limits currently in T/Ss 3/4.2.2 or 3/4.2.6, other than those related to APDMS and those discussed in the next paragraph, have been removed or reduced in our proposed revisions.

In addition to the changes described above, T/S 3.2.2 has also been revised to eliminate the need to place the reactor in Hot Standby to perform the Overpower hT trip setpoint reduction when this setpoint is AEP:NRC:0916X Attachment 1 Page 16 of 18 required to be reduced by Action Statement a.Our review of this requirement has determined that the reduction can be performed while the reactor is at power.The change in setpoint can be accomplished one channel at a time with bistables on the affected channel in the tripped configuration; therefore, there is no need to impose a transient on the reactor systems, which is inherent in changing from Nodes 1 to 3.This change is consistent with guidance provided in Draft Rev.5 of the STS.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 The changes included in this group (with the exception of the Overpowers T trip setpoint reduction) should not involve a significant increase in the probability or consequences of an accident previously evaluated.

These changes are administrative in nature and do not delete any requirements other than those associated with APDMS.As described earlier, APDMS is an option and is not required by T/Ss.For the Overpower 5 T trip setpoint reduction, the change is consistent with guidance provided by the Commission through the issuance of Draft Rev.5 to the STSs.Although the changes may increase the probability or consequences of an accident, the results should be no worse than those previously accepted by the Commission through their issuance of Draft Rev.5 to the STSs.Criterion 2 The changes other than the Overpower L T trip setpoint reduction are administrative in nature.They do not introduce any new modes of plant operation, nor do they require physical changes to the plant.The changes associated with the Overpower 5 T trip setpoint are consistent with guidance provided by the Commission through the issuance of Draft Rev.5 of the STSs and are presumed to be acceptable on that basis.Thus, we conclude that the changes will not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.

Criterion 3 The changes included in this group (other than the Overpower~T trip setpoint reduction) should not involve a significant reduction in safety margins, since they are purely administrative and in no way reduce previous requirements for safety.Changes associated with the Overpower~T trip setpoint reduction may involve reductions in safety margins, but the results of the change are clearly within limits found acceptable to the Commission through their issuance of Draft Rev.5 of the STSs.

AEP:NRC:0916I

'ttachment 1 Page 17 of 18 Lastly, we note that the Commission has provided guidance concerning the determination of significant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The first of these examples refers to changes which are purely administrative in nature: for example, to achieve consistency throughout the T/Ss, to correct an error, or to make a change in nomenclature.

The changes in this group (other than the Overpower 6 T trip setpoint reduction) are purely administrative in nature.They are intended'to improve T/S readability by eliminating the APDMS option not currently exercised, and by rearranging the T/Ss to make them easier to use.No reductions in safety requirements will occur as a result of these changes.As for the Overpower 6 T trip setpoint reduction, this change is similar to Example 6 published in the Federal Register.This example refers to changes which may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria.The elimination of the requirement to place the reactor in Hot Standby to perform the reduction does constitute a relaxation of a pr'evious requirement, but the results of the change have been found acceptable by the Commission through their issuance of Draft Rev.5 to the STSs.Based on the above, we conclude that the examples cited are applicable and that the changes involve no significant hazards consideration.

12.Changes for Consistenc With STS The twelfth group of proposed changes consist of those that are requested'o make our T/Ss more consistent with Rev.4 of the STS.These are the changes numbered 57, 58, and 96 in Attachment 10, which also includes a description of the changes.Per 10 CFR 50.92, a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not: (2)(3)involve a significant increase in the probability or consequences of an accident, previously evaluated, create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or involve a significant reduction in a margin of safety.Criterion 1 As these changes in general represent relaxation of current T/S requirements, they may involve an increase in the probability or consequences of an accident previously analyzed.The results of the changes, however, have been reviewed and found acceptable by the Commission through their issuance of Rev.4 to the STSs.Thus, we conclude that any increase in probability or consequences would not be significant.

1 4 AEP:NRC:0916I Attachment 1 Page 18 of 18 Criterion 2 As these changes will involve no physical plant changes and no T/S changes.which are not consistent with Rev.4 of the STSs, we conclude that they should not create the possibility of a new or different kind of accident.from any previously evaluated.

Criterion 3 Because these changes represent relaxation of present T/S requirements, they could potentially involve a reduction in safety margin.However, these changes are all consistent with those found acceptable by the Commission in Rev.4 of the STSs.Thus, we conclude that any reduction in margins of safety are insignificant.

Lastly, we note that the Commission has provided guidance concerning the determination of significant, hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.

The sixth example refers to changes which may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria.The changes included in this group are consistent with Rev.4 of the STSs.Although they may reduce safety requirements, the results of this change have been evaluated and found acceptable by the Commission.

Based on the above, we conclude that the example cited is applicable and that the change involves no significant hazards consideration.

Chan es to the Bases In addition to the changes to the T/Ss described above, we have also proposed changes to the Bases section to reflect both changes in the safety analyses and changes in the T/Ss.Descriptions of these changes have been included in Attachment 10.Conclusion In conclusion, we believe that the proposed changes do not involve significant hazards consideration because operation of D.C.Cook Unit 2 in accordance with these changes would not: (1)involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed, (2)create the possibility of a new or different kind of accident.from any accident previously evaluated, or (3)involve a significant reduction in a margin of safety.This conclusion is based on our evaluation of the changes, which has determined that all proposed changes which are not administrative in nature, consistent with the STS, or consistent with the design basis of the plant are clearly traceable to the Cycle 6 safety analyses, as referenced by Attachment 10.Assuming Commission acceptance of these analyses, it is our belief that they successfully demonstrate that applicable safety limits and margins to safety will be maintained.

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