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:
Shared Package
ML17334A522	List: ML17334A521 ML17334A523
References
NUDOCS 8404030229
Download: ML17334A523 (106)

Text

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 are indicated by bar in the

'hanges a

right-hand margin.

'Ihis dccun nt contains iafornation pxcpxietazy to Westinctmuse Elect~

Co~zation and An~can Elec"~ Power Sezv'ce Cozpozaticn; it is which Is fuz2LLshed ~

suhnitted in confidence ard is to he used sole+ for Nw purpose for it dccQm'.?It and such infozIGaticn Ls not to he zepzoduced, tzananitted, disclosed or used at2~vtse in whole or in par" 8404030229 820929 F

PDR ADOCK 050003l5 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 Title ~cC Zntzxduction AEP-1 System Ful~ions AEP-1 1.1.1 'Zechnical Support Center AEP 1 1.1.2 Safety Parm~xs Display System AEP-2 1.1.3 Nuclear Data Link AEP-2 1.1.4 Bypass 6 Znoperable Status AEP-3 Zr~tion. System 1.2 Eb~rt Basis

2. ~ Data Acquisition a Display System 2.1 Cat@uter System 2.2 System 2.3 Ehta Display System 2.3.1 Cnsite Technical Support Center 2.3;2 Contxol Hocm 2.3.3 Btarger~ Cpezating Facilities

3. Cnsite Technical support Center AEP-9 3e1 Desian Basis AEP-9 3.2 Znput Detexminatian AEP-10 3.3 OTSC Ccexator Zntex ace AEP-11 4 Safety Pazaratexs Display System AEP-30 4.1 Purpose AEP-30 4.2 Znput Detemunation AEP-30 4.3 Man-<~hixm Zntexface AEP-33

5. Bypass & Jr~able Status Zndication System AEP-47 571 Purpose AEP-47 5.2 AEP 47 5.3 ManW~>e Zntexface AEP-47 6.

Section Title 7.

7.1 TSC ~~ Supply Systems to the TSC Catguter AEP-56 AEP-56 7.1.1 ~ UPS System AEP-56 7.1.2 Cons~Ra of Pm'upply AEP-56 7.2 Pamr to the TSC Complex AEP-57

8. AEP-58 8.1 Task Functions Perfoznad by EnLLviduals AEP-58 in the TSC.

8.1.1 AEP-58 8.1.2 AEP-58 8.1.3 AEP 58 8.1.4 Technical Support AEP-59 8.1.5 8.2 Yanageaant ~zt AEP-59 Emergency Functions Pexfozmed in the AEP 59 TSC/ECP for each Erargency Class.

8.2.1 Chusual Event AEP-59 8.2.2 Alert AEP-60 8.2.3 Site and Genial Bmzgehcy AEP-61 8.3 Functions of Individuals Reporting AEP-62 to the ECF.

9. TSC Record and Data Availabil'ty AEP-63 9.1 Controlled Plant Specific Beferer~ i<wterial AEP-63 9.2 Chca~lled Enfozma~ and Tec.'nical AEP-64 Referer~ Ywtexial.

9.3 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.

1 1.2 ~ M~FETY STATIC DISPLAY (PSSD):

0 This PSSD systen was designed in accordance with the guidelines for This in the Safety PSSD system, ~

Paxam a faxmat that. can he te Disp1ay System (SOS )

easily xeax~ed hy +~ contxol of NGREG displays the safety status af Nm plant 0696.

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.

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s CI 2.3 DATA DZSPIAY SYSTEM 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 I

Ccaguter System. A shared video hard ccpier is provided for s

I s

the three satellite Cps.The satellite stations are arranged so that visual access from the ccrmrand station can be maintained while still providing sufficient xccm to minimize noise and distrutanoe. For printing lengttF reports, a line h

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.

2.3.3 EE Ooeratin Facilities (EOP):

hs 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 s

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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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Figurc 2.1. Tt.'chnical SUpport Complex SYstcm Configuration

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Sensor Signals non-'safety Sa ety syst.

syst., sianals sianals iso lato rs

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~Control Board I

Indication I Plant Isolator tors Process I Computer I GISX Displays X/0 Canine - ~</O Ca@inc.l IPSSD I

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TSC r Tsc I Displays I SYS E.'l e

I TECsi SUPPOFT CENTER site 8 oundary E'igu e 2.2: TSC Computer System Schematic.

TSC BXSI PSSD NUCLEAR (non- DATA iconic) 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.

2 7

~51A

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.

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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.

'=8 CAC1 1 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.

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2-9

5YZGHGHOUSE PROPRlEFARY CLASS 2 iI 4

II "5 TABLE 3.1

- CRITICAL P'LANT SYSTEMS/FUNCTIONS

,4

.Reactivity Control

,Vrimary System Inventory 5 55 5 Core Heat Removal Capabilities rI 5 Availability and Capacity of Heat Sinks hg 5 c;3

~, Containment Integrity 5

5

.5 5

- '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

• 10 AEP 14

, Table 3.2 TSC Paxarratmrs List Variables Min. No af Signals bet leg tarp 0-700 deg F

-RCS cold leg tarp 0-700 deg F

-BCS pressuxe 0-3000 psig

-Beactor water Level 0-100  %

-KS lxxcn concentxation 0-5000 pcm

-Pressurizer water Level 0-100 0

-Steam generator Level Wide xange 0-100 0 Nanna'ancp 0-100 0

-Steam Line pressure O-a4OO psig

~ntainFient pxess suxe +36 psig 589 '-599 'lev.

599 '-614 'lev.

0-100 0

~denote storacp tank level 2 0-100 0

-2oxic acid tank level o-aoo ~

-Aux feed warm flac'eed 0-250 KLbs/hr water flow

~gh ~ injection f1cw 0-5000 K1bs/hr 0-200 cpm AEP-15

"4

~ TSC Table 3.2 Paranaters List Vaziab les Min. No of Signals

-Low head injection flew 4 0-5500 gptn 16 0-2500 deg F 44 anent cooling water flow 2 0-10000 gptn

~agormnt ccoling water temp. 2 32-200 deg F 0-30  %

-Contaimnent targerature 8 0-100 deg F 4

'4

~Neutron flux 0-120  % pram

-Contml rod position 53 Pall in or rot e -Prirrary system

-Sec. syst.

relief relief

&

valves

. 4 Closed-not closed 4 Closed-not closed Closed-not closed

-P2R relief ~ pressure 1 0-100 psig O-10O S

-PZR relief ~3c 1m'. 1 50-350 deg F

-BCS degre of subcooling N/A 200 sub-5 super

-Accunulator level 0'-100 8

-Accunaxlator pressure 0-700 psig

-AcaxaQator isolation valves 4 Closed-r~ clcsed

-Aux building sump level 0-flccd level

-BHR system flow 0-7000 apn

,4

Table 3.2 TSC Paxarretars List Variables Min. No of Sicnmls

~ heat, ex. outlet temp. 0-400 deg F

~ric acid chaupir@ flaw 0-10 pe

-KS let-dawn flaw 0-200 gpn

-BCS nake-up flaw 0-200 cd

~xg

-Status af standby ~

vBDtilatich dcntKer

-Kigh radioactivity liquid closed-nat closed Emxgized 0-100 8 or not tank level

-Badiaactive gas decay tk press 4 0-150 psig

-Beactor Coolant Punps status 4 0-1200 anps

-PZR neater bank status 0-200 anps

<<Wtmrolcxy Mind dizection 0-360 deg 0-100 miles/hr Atm. delta temp. 0-50 Peg F

-Badiation 2 Car~ant area xadia~ 1 . 1-10E4 mR/hr 1 10-10E6 ~~

Containmzt air auriculate 10-10E6 axn QCit Vent radio gas 1O-1OE6 ~

Chit Vent iodine 10-10E6 cd AEP-17

Table 3.2 TSC Pazaneters List Variables Min. No. of Si ls

- Radiation (continued)

Steam gen. blow down 10-10E6 cpn Condenser air ejector . 1-10E4 mR/hr Cooling water East . 10-10E6 cpn Ccoling water West 10-10E6 cpn Service water East 10-10E6 cpn Service water West 10-10E6 cpn Waste Ziquid off-gas 10-10E6 cpn Waste gas decay 10-10E6 cpn Control rccm area . 1-10E4 mR/hr Spent fuel area .1-10E4 mR/hr ClarLzg pp room area .1-10E4 mR/hr Ncrta 1: Degree of subcooling will he independently calculated by the detectors.

TSC ccnauter.

Note 2: We radiation signals listed above are signals from the

'I existirg radiation 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 8 hhr 4~

tYES1'INAllOUSE PltOPAlEfAQ'LASS 2 TABLE 2 3 I'SC INSTRUHENT BASIS PARAHETER INITIAL EVENT DIAGNOSIS+ u BASIS .(b,c)

Containment Pressure - Determine if break is inside or outside - Honitor containment conditions of contairunent Steaml ine Press.ure - Determine if hiqh energy secondary l)ne - Haintain an adequate reactor rupture occurred heat sink

- Honitor secondary side pressure to:

.- verify operation of pressure control steam dump system

- monitor RCS cooldown rate Narrow Range Steam Generator - Determine if malfunction of secondary side - Honitor heat sink Water Level system has occurred

- Haintain steam generator water level Wide Range Steam Generator Water None - Determine if heat sink is being Level maintained Boric Acid Tank Level - None - Verify RCS boration system functions for adequate reactivity control Condensate Storage Tank Level - None - Haintain adequate water supply for auxiliary feedwater pumps Refueling Water Storage Tank None - Verify adequate supply of Leve I 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

'PfYiTIHGNONE PAOPAIHAN VMS 2 IAOLE 2-3 (Continued3 I

TSC INSTRUHENT OASIS I PARAHETER INITIAL EVE NT DIAGNOSIS* OASIS (b,c) I Wide Range Th and Tc - None - 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

ICSTIHGIIOUSE I AOPRIDARY CIASS 2 0

TABLE 2-3 (Continued)

TSC INSTRUMENT OASIS PARAMETER INITIAL EVENT DIAGNOSIS* BASIS (b,c)

System Wide Range Pressure - None - 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 - Determine NPSll for recirculation

>as occurred inside or outside containment mode cooling

- Determine which equipment in con-tainment is submerged Condenser Air Ejector Radiation Determine if steam generator tube leak - Monitor radioactivity release

. has occurred path to environment Steam Generator Blowdown Radiation - Determine if steam generator tube leak - Monitor radioactivity release has occurred path to environment Contaienent Radiation Determine if high energy line break or fuel - Moni tor radi oactivi ty release mishandling accident 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

- Honitor environmental conditions around equipment in containment Auxiliary Feedwater Flow Hone - Determine if sufficient flow exists to maintain heat sink Iligh tlead Safety Injection Flow - Hone - Determine that ECCS is deliyer-ing flow

- Honitor ability to keep core covered Low llead Safety Injection Flow - Hone - Determine that ECCS is deliver-pl&

ing flow I

hJ h)

- Honitor ability to keep core covered

- Infer spray operation Area Radiation Honltoring in -'etermine if source of accident is outside', - Honitor accessibility to plant Auxiliary Building and Control cont a I runent bu i l ding zones/equipment Room

- 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

gf.'f tttQIIOUSE I AOPNITARY CLASS g TABLE 2-3 (Continued)

[

TSB IRSTRBMBRT BASIS II A PARAMETER INITIAL EVENT DIAG1IOSIS+ BASIS (b,c)

- Honitor habitability of the control room 7

Core Exit Thermocouples - None Determine if core is being cooled I

Neutron I'lux - None - Monitor ability of reactivity control systems to keep the core subcritical I Determine if plant is in a safe shutdown condition Degree of SuLcooling of - None Haintain adequate reactor heat Primary Coolant sink

- Haintain safe shutdown condi-tions I

Primary System Safety and - None - Haintain primary system inventory Relief Valve Position

- Monitor radioactivity release paths into the containment Pressurizer Relief Tank - None - Monitor capacity to relieve Pressure, Temperature, and Level primary coolant pressure

- Monitor radioactivity release paths into the containment Containment Isolatton Valve - None - Monitor radioactivity release Position 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

0 WESTINGIIOUSE PROPRIETARY CUSS 2 TA""E ~. 3 (Continued)

TSC INSTRUHENT BASIS PAR AHETER INITIAL EVENT DIAGNOSIS* BASIS (b,c)

)

Secondary Safety, Reliefs, - Hone - Honitor. radioactivity release and Atmospheric Dump Valves paths to the environment

- Honitor secondary system integrity Accumulator Tank Level - None - Honitor primary system inventory

- Determine whether the accumulator tanks have in5ected into the RCS Accumulator Isolation Valve - None - Determine system operation Position RllR System Fits - None - Hopitor primary system inventory

- Honitor core heat removal capabilities RllR Ileat Exchanger Outlet - None - Honitor core heat removal Temperature capabilities Component Cooling Mater Flw - None - Honitor system operation of and Temperature a critical support system

• Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.

SESTINGIIOUSE PROMJETAB'LASS 2

551A

WESIINGIIOIISE PROPRIETARY CQSS g TABLE 3 3 (Continued3 TSC IHSTRUHfHT OASIS PARAHfTER INITIAL fVftIT DIAGtlOSIS* BASIS (b,c)

Boric Acid Charging Flow - tlone - Honitor pr imary system inventory

- Determine boron concentration for reactivity control

- Honitor ability to control RCS pressure Letdown flow - None or pri~~ry system inve,,or

- Honitor ability to contr ol RCS pressure

- Honitor core heat removal capab ili ty

- Determine boron concentration for reactivity control Water Level - ttone - Honitor environmental conditions in Closed Spaces Around Safety around required safety equipment fquipmcnt )n Auxiliary Building outside of containment Emergency Ventilation Damper - Hone - Ensure proper ventilation to Position vital areas under post-accident conditions ltigh Level Radioactive Liquid ,

- tlone - Honitor capacity to contain Tank Level and store radioactive liquids I

'Certain indications on this table are used as secondary diagnoses as the operator proceeds through Post-Incident Recovery.

5251A gfg]NIIOIISE P.".".0" IQ'ARY Clh.S 2

~

~

Vlf."TltlGllOUSE PROPAIETAAY CLASS 2 TABLE 2-3 (Continued)

TSC INSTRUMENT BASIS PARAMETER INITIAL EVENT DIAGNOSIS* BASIS (b,c)

Radioactive Gas lloldup Tank - tlone - Honitor capacity to contain Pressure and stare radiaactive gases of All Electric

'tatus Power - None - Ensure adequate electric power Supplies and Systems ta safety and suppart systems Effluent Radioactivity Noble - Honitor radioactivity release Gases, Radiohalogens, and paths to the enviranment Particulates Plant and Environs Radioactivity - None - Monitor release of radioactive (Permanent and Portable materials not covered by Instruments) effluent monitors Sampling System - tlone - Oetermine RCS chemistry for reactivity control and extent of fuel clad damage Meteorology (wind speed and - None - Monitor radioactive effluent direction temperature prof lie, transportation for emergency and precipitation) planning, dose assessments, and source estimates Containment Atmosphere temperature - None - Monitor containment integrity and ttydrogen Concentration

- 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.

CLASS 2 VIESTlrlcttOUSE PROP;;tETNW 5251A

iNgiiNGHOUSi PROPRIEMRY CLASS 2 Systems Status - Reactor Coolant System Loop 1 Loop 2 Loop 3 Loop 4 T average ('F) 595.2 595,2 595.2 595.2 Overpower DT PoPWR) 110.0 110.0 110,0 110.0 Overtemp. DT (%PWR) 110.0 110.0 110.0 110.0 Cold leg temp. (narrow range) ('F) 559.8 559.8 559.8 559.8 Hot leg temp. (narrow range) )'F) 624.0 624.0 624.0 624.0 Reactor coolant flow (%) 100 0 'G~ 0 100.0 1GO.O Reac'.or coolant pressure - WR (PSlG) 2250.0 2250.0 2250.0 2250.0 Pressurizer pressure (PSlA) 2250.0 Pressurizer vapor temp. (') 563.8 Pressurizer liquid temp. ('F) 565.2 Pressurizer relief tank pr.ssure (PSlG) 1.5 Pressurizer relief tank level ('h) 77.6 Pressurizer relief tank temp. ('F) 110.3 Pressurizer safety relief temp. ('F) 120.0 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

. TO406 RCS Loop 1 Cold Leg T 5472 0:700 OEGF . Normal PO480 RCS Pressure 2234.1 OOOO PSlG Normal LO421 Stm Gen 2 Narrow Range Level 39.1 0:100 PC Low PO549 Steamline Pressure 893.0 0:1100 PSlG Normal LO103 RWSi Level 100.0 0;100 PC Normal LO114 Boric Acid Tank Level 98.8 0:100 PC Normal LO119 Condensate Storage Tank Level 58.4 0:100 PC Normal LO947 Containment Bldg.'Vater Level 3.3 0:160 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 (b,c,e) 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:

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 (b,c,e) associated with the two potential event typegs the PSSO incorpar ates Ltwo !b,c,e) cperating mades. The ',ir't made (TERMIRATE MODE) is itive whi1e:hge 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 t

for each of the operating mode/a 1s as foll'ows:

(b c e) cEMIMATE MODE l

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.

MITIf)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

(VESTNGHQUSE PRQPRlETARy CLASS 2 LB. Radioactive Release (b,c,e)

5. Containment Environmengt By addressing Lacy safety concerns, the consequences of abnormal events (b,c,e) can be limited or mitigatgd, tThe tey safety concerns can be related to specific abnormal occur- (b,c,e)-

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 (b,c,e) fo 1 1 ows:.

Ll. The inputs selected must represent a minimum sat sufficient for (b,c,e) monitoring all possible events including those which might not have been anticipated.

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 (b,c,e) 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 4-3 "43"8 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 The PSSO system will- process the defined input data set of plant param<<,

(a,b,c) stere atftwo second interval/sand generate displays for redundant PSSD (a,c) 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!.

4 a Nay. " 4 Jvsaa's 'w V'-"S IflGHGUSE PROPS!EERY CLASS 2 Activity: Reacti on (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:

Goal:

Terminate/Mitigate 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 (a,c,f) 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 4 g r,3 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 4-6 AEP-35 5435A

)';-ST!,'su,",OUSE PROPrltTAnY CLASS 2 The'inforaation atfLeve1 2 is an expansion of each of the key safety (a,c,f) 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@

4~7 543""A AZP-3 6

• e 'e e j

ÃESTlHQHOUSE P!OPRfET'qY Ct ~SS Z TASLE 4-1 PLANT SAFETY STATUS..OISPLAY>> SAFETY GOALS - TERMINATE MOOE TRANSIENTS (b,c,e)

Reactor Control Systems Malfunction ee e4 e

Stop rod motion

.'*C

'~1

-+i C- Maintain core thermal and nuclear parameters within limits A

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 Maintain pressurizer pressure and level eg

'e 0 4-8 AEP-37 54351

V/EST1HGHOUSE PRCPRluARY CLASS 2 TABLE 4-2 PLANT SAFETY STATUS OISPLAY - SAFETY GOALS - MITIGATE MOOE TRANSIENTS (b,c,e)

Reactor Trip 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 4-9 AEP-38 5435A

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 VV 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 Coolant Reactor Reactor Control 'akeup Coolant System Control System Inadvertent System Malfunction Malfunction Oepressurization Leak avg X X ref X X Rod position X X Oelta T X Startup rate X Count rate X Pzr. pressur 0

Charging flow X Pzr. level X Comp. cool X H20 rad Containment rad X Air eject rad. X Blowdown r ad. .X Cont. humidity X Cont. temperatures X Cont. oressure X Prz. discharge X piping temps PRT pressure X X PRT level X X PRT temps X X RCP seal tempera- X ture RCP seal flow RCP seal level YCT flow 4-11.

jJ imp AEP-4 0

Y~wRCHOVSE PRCPRIETAC CLASS 7 PLANT SAPPY STATUS OISPLAY "IlTTGATE WOE PARAvETERS VWable Trans1cnt Steam Operation loss of Loss af Generator Reactor Station Gnergency with 'latural Ceo lant Secondary Tube Trip Blackout Bar'ation Clrcul atlon Accident Coolant Ruature.

Reactar trip breaker X Startup rate X Reutran flux X X Rod pos1tfon X X Turbine tHp X Blackout signal X Tavg ( thermacoup 1 es ) X Rad bottom 3nd. X Primary pressure X Stcam flex X X Feed flow X X Pressurf ter level X X X Care thcrmacoup les X X X Cont. radar at<an X Afr Qectar rad$ at5an X Slowdown radiation X

Cont. pressure X X" Pri. M.R. tanp. X X Stcam pressure I ~

X X Cant. sump level X Cant. temperature-Cant. huahdl ty Charging flaw X S.G. level X X B.A. tank levei X Aux. fmd flaw X Sa 'law X RMST level X X CST level X X a-lZ.

AEP-4 1

)cSBA

16708-1 ygggtfGHOUSE PROPRtH'ARY CLOS 2 (a,c)

CONDITIONS NORMAL DETECTION

?

YES SAP ETY IMPLEMENT IDENTIFY LIMITS REACTION RESPONSE RESPONSE VIOLATED NO DIAGNOSE PROBLEM DIAGNOSIS TAKE COR R ECTIV'E TERMINATE ACTION /MITIGATE Figure 4-1. Operator Response Madel AEP-42

16643 10 WESTINGHOUSE PROPRIETARY, CLASS 2 T p Qrepftie, LEVEL 1 OISP LAY Olsplay C'y X Loop, TSAT Reactor TAVQ TH TC Coolant PR. Press Inventory STIjf/FO Flow Przr Level P Steam RCP's Cte. Leatown

, LEVEL 2 OISP'LAYS Pressure Relief Vlv. W.R. TH. TC.

Safety Vlv. Core TC's Spray TSAT. Etc.

Heaters PER. TCS Sensors, Comparisons LEVEL 3 of Redurufant OISPLAYS IVleasurernent Error Ctteeks j Inputs Figure 4-2. Display Structure of Plant Status Display AEP-43

WESTlHGHOUSE PROPRlETARY CLASS 2 Primary Tavg (Value) F Startup Pressurizer Rate Pressure (Value) Oec/Min (Value) psi

/

/

/

/ Containment Pressurizer /

Level Humid (Value) o/o (Value) 4k / Temp (Value) 'F

/

/

/

Charging Radiation Flow Contmt (Value) GPM Blwdn Air Eject Steam Gen Level (Value) Io Fig"~ 4-3. S~pie Display plant Safety St tus Dl~tay AEp-44

NESTINGHOUSE PROPRIETARY CLASS 2 Primary T avg (Value) 'F Pressurizer Startup Pressure Rate (Value) psi (Value) Dec/Min

- ~

k 4

Pressurizer Containment Level Humid (Value) '6 (Value) % I I 'emp (Value)'F I

I I

II Charging Radiation Flow Contmt (Value) GPM Btwdn Steam Gen Air Eject Level (Value) %

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.

(a,c,f)

Temp (Value) oF (Value) 'F Tsat Startup Rate RCS W.R.

Pressure (Value) Dec/Min (Value) psi 4

~ '\

II

\

\

I I \

I Pressurizer II Containment Level Pressure I (Value) psi (Value) %

II I I

II rr

~r R. V. Radiation Level Steam Gen Contmt (Value) % Level Bid dn (Value) % 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 the bypassed or inoperable condition. The ontrol room operator must tionss of 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-AEP 48

WESTltsGHQUSE PI'.OPHl~iARY CL".SS g

/The timing i'unction enab'les the control rom aperator ta set up a count- (a,c 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.

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

~Sstem Comoonents Status

('.c)

Emergency core cooling Yalves Open/Shut

=

Pumps Operable

~

.Process High/L'o~, etc.

(level, pressure)

~ t$

y,'I

.Auxiliary feedwater Valves Open/Shut

~'a o

"a . Pumps . Operable

. Process Nigh/Low, etc.

Containment Valves Open/Shut Pumps Oper able Process =

High/Low, etc 0 spray'ontainment i so ation 1 Valves Open/Shut Auxiliary power system Breakers Open/Closed/Out

'enerators Operab.l e Voltages High/Low

'Containment ventilation Yalves Open/Shut Motors Operable Containment hydrogen Valves Open/Shu recombiners Motors Operable Component coo1ing Valves Open/Shut Pumps Oper able Service water Yalves Open/Shut Pumps 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 Steaml inc isolation

~

s4 4

Steam dump interlock Steam generator blowdown isolation 3-5 AEP-51 5251A

IESTINGHOUSF. PROPRIETARY CLASS 2 Systems Level Status cCCS Hl Head Sl Prfrnary Qteplay Accumulators Operator Etc.

Inputs Safety Continent lOtnersl lntecuon Spray Pump 1 Ready Pump 2 Out Stthsystern Cont ponent Valve t Open Level Qtapksy Containment Spray Suc pot Comoonent Cooling Support Systornl ESP Power Component Lovel Pisplay Etc.

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 Operable Intermediate Head Sl Operable Low Head SI Operable Accumulators Operable Auxiliary Feedwat r Operable Containment Isolation Operable Containment Spray Inoperable - Train A Component Contaioment Ventilation Operable Safeguards Power Source Operable Figure. 5. 2Primary Disofay Bypaued and inoperable Status indication AZP 53

WESTINGHOUSE PROPRlETARY CLAS) g CONTAINMENTSPPAy Train A Train 8 Train C VLY101 VLV201 VLV301 Pump A Suet Open Pump 8 Suet Open Pump C Suet Open YLV111 VLV21 1 VLV31 1 NAOH Supply Open NAOH Supply Open NAOH Supply Open Pump A Operable Pump 8 Operable Pump C Operable VLY102 VLV202 VLV302 Pump A Outlet Closed Pump 8 Outlet Open Pump C Outlet Open VLY103 VLV203 VLV303 Headr A Outlet Closed Headr 8 Outlet Closed Headr C Outlet Closed VLV121 VLV221 VLY321 Recirc A Closed Recirc 8 Closed Recirc C Closed Refueling Water Storage Tank LS1 CO Level Normal LS101 Level Normal LS1 02 Levei Normal LS103 Level Normal NAOH Spray Additive LS200 Level Normal TS200 Temp Normal LS201 Level Normal TS201 Temp Normal LS202 Level 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 the battezy consists of xedundar&

invextexs, and static txansfer switches.

charger converts AC to CC

~

battezy chairs,. battezy, static and ruxmal conditions, 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 INOEPENOENT INOEP ENOENT 600 VOLT BUS 600 VOLT BUS 600 VOLT BUS O'C.C. M.C.C.

BREAKER BREAKER 225A 225A M.C,C. AUTOMATIC BREAKER TRANSFER 225A SWITCH 260A 600 ~700 I

A~MP 700 AMP

'5KVA BATTERY CHARGER BATTERY 120 I CHARGER (ALTERNATE}

BATTERY 927A 40KVA 40KVA INVERTER INVERTER STATIC STATIC SNITCH SWITCH FIGURE 7. I UNIT W I UNIT W2 TSC TSC COMPUTER 8 COMPUTER 8 P 'ERIPHERALS 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.

AEP-58 L

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C

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

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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 0.

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~

105 in Attachment 10.

60~ 62~ 69'4'1~

These changes 83~ 84'8'" 90'3J 94'7/

are proposed to enhance the 98'nd 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:

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.

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 T/S 3 '.1.1 b T/Ss 3.4.1.2 and 3.4.1.3 c 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:

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.

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 0. 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 0.

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:

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.

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:

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 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:

involve a significant increase in the probability or consequences of an accident previously evaluated, (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.

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 H specification (3/4.2.3) to the DNB specification (3/4.2.5.1) for consistency with Unit 1 specifications.) This exemption is required 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 0.

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, 17, 18, 19, 23, 28, 34, 40, 42, 47, 48, 49, 54, 76, 78, 79, and 101 in 14'5, 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 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 hH ). These changes are numbered 41, 42, 43, 48 in Attachment 10. In the present T/Ss, RCS flow rate and F may be "traded off" against one another (i.e., a lower measured RCS flow rate is acceptable provided F hH 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 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 T/S values for flow are included in Category fH 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 AH) it could be anticipated that an increase, rather than decrease, in the margin to DNB under accident conditions might actually 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

ave low-low bistables. These 0 bistables are calibrated to trip when the 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 Tave channels above the reset point, 0

which is greater than 541 F, the manual. block of Low Steamline Pressure 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 0 F and below 540 0 F. If this description is read it literally, could be inferred that P-12 is established when o Tave is greater than or equal to 544 0 F and when Tave is less than 540 F.

Additionally, the manual block 0 of safety xn3ection actuation would0 not be permitted until below 540 F, when in fact the 0

setpoint is 541 F. We propose to rewrite P-12 in terms of the 541 F setpoint, which is similar to 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 Tave . This particular Safety Injection may be blocked if the P-12 Low Low ave f.'ignal is present. High steam line flow it coincident with P-12 Low Low T ave does not provide a Safety Injection; 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:

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.

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 it was decided not to operate with APDMS. Therefore, we are delete T/S 3/4.3.3.7, and to revise T/Ss 3/4.2.2 (F (Z)) and

'easons, proposing to 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 Qdefinition of APL in%he proposed T/S 3.2.6. No 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:

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.

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:

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.

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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