From: Richard WebsterTo: Lamb, JohnCc: Janet TauroSubject: FW: Oyster Creek 2.206 Petition - PRB Immediate Action DecisionDate: Thursday, January 03, 2013 5:24:44 PMAttachments: 19890900-oc-sourcebook.pdf

John: Please see e-mail below from a nuclear expert.  Could you clarify whether the ESWS is designed towithstand the maximum flood level? Thanks  Hello Richard: Fukushima demonstrated that reactors could survive seconds, maybe even minutes, withoutcooling water. Seriously, there are more cooling water systems at Oyster Creek than the circulating waterand service water systems. Attached is the NRC's sourcebook for Oyster Creek from September 1989. On the 13th page of this 110 page document is a schematic of the cooling water systems atthe plant. The circulating water system (CWS) and the service water system (SWS) are non-safety-related systems used to cool equipment and structures needed for the day-to-day business ofgenerating electricity. The emergency service water system (ESWS) and the fire protection system (FPS) arestandby systems that are credited in the safety studies for cooling equipment and structuresneeded to mitigate accidents. In theory, reactor cores can be adequately cooled even if the SWS and CWS are totallyunavailable. In practice, SWS is often used to refer to both the SWS and the ESWS, since some of thepiping and valves are shared between the systems. In the text you sent, someone referred to the service water system. Assuming that it is just theSWS and does not refer to the ESWS, its loss can be okay.

From: Richard Webster [rwebster@publicjustice.net]Sent: Thursday, January 03, 2013 3:03 PMTo: pgunter@beyondnuclear.org; Dave Lochbaum; Arnie Gundersen; Jim RiccioCc: Janet TauroSubject: FW: Oyster Creek 2.206 Petition - PRB Immediate Action Decision

mailto:rwebster@publicjustice.netmailto:John.Lamb@nrc.govmailto:janettauro007@gmail.commailto:pgunter@beyondnuclear.org

I I .' , , ,

NUCLEAR POWER PLANT SYSTEM SOURCE

.. . '.., ,,..,.l,ll

OYSTER CREEK I

Oysrer Creek

I 1 O F CONTENTS

p&,J

s utxm OON PLANT'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I

ICA7'10N OF SIMILAR KUCLEAK POWER PLASVI'S . . . I

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I \ ~ e r Creek . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

A I ' I I ' A l k l i ~ i r r ~ o n of Syn~bols Used in r t ~ r System and I . ayou~s 1)rawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

, , , . . Oyster Creek

. . . . . : . , ..

LIST OF FIGUR,ES (continued) . . . . , , , I , , . .

3.10-1 Oyster Creek Reactor Shutdown Cooling System ......................... 3.10-2 Oyster Creek Reactor Shutdown Cooling System Showing Component ....................................................................... Locations 4- 1 General View of Oyster Creek ~uc lear Power Plant and

........................................................................ Vicinity 39 ............................................ ster Creek General Site Plan.. 40 , :,

. . . . . . . ,.. ...................... Building Elevation Drawing.. 4 1 .................... Oyster Creek Reactor Building - Elevation (-) 19'-6" 42

.... 4-5 Oyster Creek Reactor Building - Elevation (-) 6-5" and (-) 1'-11". 43 ........................ 4-6 Oyster Creek Reactor Building - Elevation 23'-6" 44

4-7 Oy5,er Creek Reactor Building - Elevations 33'-5" to 38'-0" ............ 4-8 Oyster Creek Reactor Building -,Elevation . . 51'-3" ........................ . . 45 4-9 Oyster Creek Reactor ~ u i l d i n ~ - Elevation 75'-3" ........................ 46 4-10 Oyster Creek Reactor Building - Elevation 95'-3" ........................ 47 4-1 1 Oystcr Creek Reactor Building - Elevation 119-3" ....................... 48 4-12 Oyster Creek Turbine Building - Elevations 0 ' -0 and 3'-6 ............ 49 4-13 Oyster Creek Turbine and Office Buildings - Elevations 23'4

' , .......................................................................... to 27 '0 50 4- 14 Oyster Creek Turbine and Office Buildings - Elevation 35'-0.. ........ 5 1 4-15 Oyster Creek Turbine and Office Buildings - Elevation 46-6". ......... 52 4-16 Oyster Creek Turbine and Office Buildings - Roof Elevations ..........

- ........................... 4-17 Oyster Creek Intake Structure Elevation 6-0 53 4-18 Oyster Creek Freshwatcr Pumphouse - Elevation 1 2 - 0 ................ 4-19 Oystcr Creek Emergency Diesel Generator Building, Elevation 23'0 . 54

Oyster Creek

LIST O F FIGURES (continued)

E l u c e;ire

A-1 Key to Symbols in Fluid System Drawings ................................

A-2 Key to Symbols in Elecmcal System Drawings ...........................

A-3 Key to Symbols in Facility' Layout Drawings ..................... ....

Oyster Creek

LIST OF TABLES

............. 3- 1 Summary of Oyster Creek Systems Covered in this Repon

3.1 - 1 Oyster Creek Reactor Coolant System Data Summary for Selected Components ........................... ... .......................................

3.2-1 Oyster Creek Isolation Condenser System Data Summary for Selected Components ......... .... ..................................................

3.3-1 Oyster Creek Emergency Core Cooling System Data Summary for Selected ......................................................................

, , , , , .,., . . , , . .,.. ?,,a:.,L,,.i. , ........................

ontainment~~ray System Data Summary for Selected ' , , .. ' ...................................................................

,: iii oyster Creek Elecujc Power , System I t 1 . . Data Summary forselected

, . C o m p o n ~ n t s ....................................................................... , ,,

, , , , , . , ' ,

........ Pamid ~ i ' s t i n ~ of Electrical ~,q&es and Loads at Oyster Creek..

Oyster Creek Conuol Rod Drive Hydraulic System Data Summary for Selected Components.. ................................................... Oyster creek Fire Protection System Data Summary for Selected Components .... ; ............ .., ...............................................

, , ,

Oyster Creek Emergency Service Water System Data Summary / I for Selected Components. ....................................................

Definition of Oyster Creek Building and Location Codes ................ Panid Listing of Components by Location at Oyster Creek .............

Component Type Codes.. ....................................................

Oyster Creek 1 , . . . I

. . . .. ,. . . .

The information in this repon has been developed over an extended period of time based on a site visit, the Final Safety Analysis Report, system and layout drawings, and other published information. To the best of our

ge, it accurately reflects the plant configurati the ion was obtained, however, the information in has independently verifiedB'*:Ule licensee or the NR

This sourcebook will be periodically updated with new andlor replacement pages as appropriate to incorporate additional information on this reactor plant. Technical errors in this should be brought, t

wing:

Mr. Mark Rubin U.S: Nuclear Regulatory Commission Office of Nuclear Reac~or Regulation

Division of Engineering and Systems Technology Mail stop 7E4

Washington, D.C. 20555 ,

With copy to: I /

Mr. Peter Lobner Manager, Systems Engineering Division

Science Applications International Corporation 10210 Campus Point Drive

San Diego, CA 92131 (619) 458-2673

Correction and other recommended changes should be submined in the form of marked up copies of the affected text, tables or figures. Supporting .

,

documentation should be included if possible.

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: Oyster Creek . , . , .

OYSTER CREEK SYSTEM SOURCEBOOK

This sourcebook contains summary information on the Oyster Creek nuclear power plant. Sumnlary data on this plant are presented in Sectio;~ 1, and similar nuclear power plants are identified in Section 2. Information on selected reactor plant systems is presented in Section 3, and the site and building layout is illustrated in Section 4. A hihliography of repons that describe features of this plant or site is presented in Section 5. Symbols used in the system and layout drawings are defined in Appendix A. Terms used in data tables are defined in Appendix B.

1 . SUMMARY DATA ON PLANT Basic infom~ation on the Oyster Creek nuclear power plant is listed below:

- Docket number 50-219 y Cenaal Power and Ligh!. ,., , .,,:,,

, . . . . .; . , - Jersey. 2 miles south of Forlted'River - Reactor type BWR - NSSS vendor General Elecmc - Power ( h l W W e ) 1930/670 - Architect-engineer Bums and Roe - Containment type Mark I

2 . : IDENTIFICATION O F SIMILAR NUCLEAR POWER PLANTS The Oyster Creek plan1 has a General Electric B W 2 nuclear steam supply

system (NSSS), isolation condensers, and a Mark I type containment. Other BWRs with isolation cpndcnsers arc Nine Mile Point-1 (RWR/2), Dresden 2 and 3 (BWW3). and Millstorp . . l , (BWRn).

,;Oyster Creek employs different systems for core coolant injection and shutdown cpoling than most BWRs.

, ,

3 . SYSTEM INFORMATION This section contains descriptions of selected systems at Oyster Creek in terms

of gener,a!,:function, operation, system success criteria, major components, and suppon system requirements. A summary of major systems at Oyster Creek is presented in Table 3- 1. In th&"Repon Section" column of this table, a section reference (i.e. 3.1, 3.2. etc.) is provided for all systems that are described in this rtpon. An entry of "X" in this column means that the system is not described in this rcpon. In the "FSAR Section Reference" column, a cross-reference is provided to the section of the Final Safety Analysis Repon where additional information on each system can be found. Other sources of information on this plant arc identified in the bibliography in Section 5.

Several cooling water systems a y identified in Table 3-1. The functional relationships that exist among cooling warn systems required for safe shutdown are shown in Figure 3-1. Details on the individual cooling water systems are provided in the rcpon sections identified in Table 3-1.

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Table 3-1. Summary of Oyster Creek Systems Covered in this Report

. . . . . . . Report .~ . . Seclion

. ~ ~

. . ~ ~

Reactor Heat Removal Systems . . - Reactor Coolant System (RCS) Same 3. 1~

6.3.1.1, 6 . 3 . 2 -1 - : - - ' I . ~ - Isolation Condenser System (ICS) Same 3.2

- Reactor Core Isolation Cooling (RCIC) Systems

- Emergency Core Cooling Systems (ECCS) : ~ i ~ h k s s u r e Injection

& Recirculation - Low-pressure Injection

N & Recirculation - Automtic Depressurization

System (ADS)

- Decay Heat Removal (DHR) System (Residual Heat Removal (RHR) System)

- Main Steam and Power Convenion .

. . . ~. , . .- . . Systems .....

. - m r Heat Removal Systems

None

Same

None

Core Spray (CRS) System.

Same

Shutdown Cooling (SCS) System

Main Steam Supply System, Condensate and Feedwater System, Circulating Water Sysrem

Head Cooling System

Table 3-1. Summary of Oyster Creek Systems Covered in this Report (Continued)

- . Reactor Coolant Inventory Control Systems - Reactor Water Cleanup (RWCU) Same

System

- 332s - See above 4.6.1,2 . . . . ,..:. . .. .. . % ~ . . . . . .. . ~~ . . , - Control Rod h v e Hydraulic System (CRDllS) Same . 3 ;7 .

~- ..- ~. Containment Systems

. .

- Primary Containment Containment Suucture 6.2.1 (drywell and pressure

~ ~ suppression chamber) -. . ~ .~ ~~. ~ -- w Secondary Containment Same 6.1.3 . ,

- Standby Gas Treatment System (SGTS) Same 6.2.3

- Containment Heat Removal Systems - Containment Spray System Same (also cools torus) 3.4 6.2.2

Containment Fan Cooler System Reactor Building Hating and X 9.4.2 Ventilating Systems

- Containment Normal Ventilation Systems Reactor Building Heating a" 9.4.2 Vennlating Systems

~ .~ . . ~. , ,

- Combustible Gas Conml Systems x; 6.2.5.2.1 '- . .. . Containment Inerring ~ ~ s t e & , ~ ~ HydrogenIOxygen Monitoring . - -~ X: 0.2.5.2.2 System

Table 3-1. Summary of Oyster Creek Systems Covered in this Report (Continued)

. ~. ~ ... ~~ . Generic Plant-Specific ~ & o r t UFSAR-Section

Svslem Name $stem Name S e t ion R f . , . . - 1 .. .. . . ~ - ... . . .

Reactor and Reactivity Control Systems - ReaaorCm Same'

~ ~

4

- Control R d System Control Rod Drive Mechanisms X 4.6.1.1, 4.6.2

- Chemical Poison System Standby Liquid Control System (SLCS)

Instrumentation & Control (I&C) Systems - Reactor Prorection System (RPS) Rwctor Trip System 3.5 7.2

- Engineered Safety Feature Actuation >

P System (ESFAS)

. .

7.3 Engineered Safety Feature Systems 3.5:, . . . .> ,~ . . ~ ~

. .

- Remote Shutdown System Local control panels 3.5 7.5.2.4.2 - Other I&C Systems Various other systems X 7.4 to 7.7

Support Systems - Class 1E Electric Power System Same 3.6 8.3

Same 3.6 8.2, 8.3 - Non-Class IE Elecmc Power System . .

. . - Diesel Generator Auxiliary Systems Same 3.6, - . ~ . ~ 8.3.1.1 .S, 9.5c4-to - - , - . 9.5.9

xi . .. Reactor FJuilding Closed 9.2.2.2 ~ :. ' . .. . . ~ .. .- - Component Cooling Water (CCW) . . . System Cooling Water (RBCCW) System

Table 3-1. Summary of Oyster Creek Systems Covered in this Report (Continued)

Plant-Specific ~ e ~ o r t UFSAR Sect- ion System Name sC

CONDENSER. - .

w

Figure 3-1. Cooling Water Systems Functional Diagram for Oyster Creek

Oyster Creek

3 . 1 REACTOR COOLANT SYSTEhl (KCS)

3 .1 .1 , . : . . . . , , . m d the Nuclear Soam Supply System (NSSS). is responsible

for directing the steam roduced in the reactor to the iuibine where i t is used to.rotare a generator and produce efecuicity. The RCS pressure boundary also establishes a boundary against the unconuolled release of ridioactive material from the reactor core and primary coolant.

1

3.1 .2 The RCS includes: (a) the reactor vessel, (b) five recirculation loops. ( c )

recirculation pumps, (d) relief valves, (e) safety valves, and (f) connected piping out to a suitable isolation valve boundary. Simplified diagrams of the RCS and imponant system i~~erfaces.ar:e,,shown in Figure 3.1-1. and,3..1-2. A summary of data on, selected RCS components, ispresqnted in Table 3.1-1.:;. .::.!: *I- ., . . , . .. , . . ~ > , ! . , , : ,

, , 3.1.3 , , . , , , . , :

During power operation, cikulation in the RCS is maintained by one recirculation pump in each .of the five recirculation loops. The steam that is produced by the reactor is piped to, the turbine via the two main steam lines. Condensate from the turbine is returned to the RCS as feedwater.

Following a uiinsient that involyes the loss of the normal heat transfer path, heat transfer from the RCS is accomplished by: (a) emergency condensers which vent directly to atmosphere (see Section 3.2), or (b) by opening the automatic depressurization system (ADS) valves and dumping heat into the containment (see Section 3.3). In the latter case. and in the case of a LOCA, heat is transferred from the containment to the ultimate heat sink by the containment spray systern (see Section 3.4), and a low pressure injection system provides RCS makeup. Actuation systems provide for automatic closure of the main steam isolation valves.(hlT,TVs) and isolation of o$er lines connected to the RCS.

3.1.4 The RCS success criteria can be described in terms of LOCA and translent

mmgation, as follows:

- An unmitigatible L O ~ A 1s not initiated. - If a rnitigatible LOCA is ininated, then LOCA mitigating systems are successful. - If a transient is initiated, then either: - RCS integrity is maintained and transient mitigating systems arc successful. or , - RCS integrity is not maintained, leading to a LOCA-like condition (i.e. stuck-open safety or relief valve, reactor coolant pump seal failure), and LOCA mitigating systems are successful.

A. RCS 1. Volume: unknown 2. Normal operating pressure: 1OOO psig at 4 7 ' F

Oyster Creek

D . Safety Valves ( 1 6) I . Set pressure: 1212 to 1239 psig 2. Relief c;~pacity: 611,000 Ibhr each

C . Power-Operated Kel~ef Valves (5) 1 . Set pressure: unknown 2. Rclicf capacity: 600.000 lbhr each 3 . Type: Electrornatic

D. Recirculation Pumps ( 5 ) 1 . Rated flow: 33.000 D m @ 120 ft. head (52 psid) 2 . Type: Venical cenmfugal

, , ' . , . ,',, . A. Motive Power < , . ,', . 1 . _ i , , . . , . _ , . . , . . .

: 7,'fie recircula(ion'puni 'ed from Non-Clss , lE sw~tchgear. ' 8 ,: ,

. ,MSIV Operqting Power , .The insirument air system suppons normal operation of the MSIVs. Valve operatio? iscontrolled by' an AC and a DC solenoid pilot valve, Both solenoid valves r?ust,be deenergized to cause MSlV closure. This design prevents 'spurious:closure of an MSlV if a single solenoid valve should fail. MSIVs are designedto fail closed if insvument air is lost o r if both AC and DC control power islost to,the solenoid pilot valves. This is achieved by a local dedicated air accumul~tor'for each MSlV and an independent valve closing spring.

' C . Recirculation Pump Cooling The reactor building closed cooling water system provides cooling water to the re~irculation pump coolers.

I

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Figure 3.1-2. Oyster Creek Reactor Coolant System Showing Component Location

Table 3.1-1. Oyster Creek Reactor Coolant System Data Summary for Selected Com~onents

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. .. . ' Oyster Creek '. , , . , . , .. /' . , ,

3 . i ' ' ' .I.SOILATi,OS ; , ' , 1 , , ' . , C ( ~ N I ) F : N S E @ ? S Y S T E ~ I ~ , . . . , (ICS) . .

, , , . , ~ , , f ' , i < ' ' ' , , . . . :id , , . ( ' . - . , , . , . . , *s%ci-

' ' l'he 1CS (also called the Emergency Condenser System) provides a backup heat sink for reactor heat when the main condenser is unavailable. The system provides a natural circulation heat transfer path for the RCS, with heat transferred to the environment by boiling secondxy-side water in the isolation condensers and venting steam directly to atmosphere. Because of its role in emergency cooling, the ICS is considered pan of the Emergency Core Cooling System (ECCS, see Section 3.3).

3 . 2 . 2 The ICS cons~sts of two full-caoacitv isolation condensers. each with a closed- ~ ~ ~~

loop flow path to and f romthr reactor a n i a v h to atmosphere. he inlet piping to the isolation condensers each contain two normally open AC-powered motor-operated isolation vnlves. The outlet piping cont;i::ls one p o ~ a l l y closed DC-powered h10V and one norrnallv , .,,.. , . - , 0pen:P.C-powered 510\'. Tn;,$$g,,>of DC-powered isol$!.i:b"..yalyis ensures succes~fu'l:ope?~tion'of t 'he" '~ys~'em 'if there is a loss of AC power. Makeup to the condensers isprovided by the Firr' I'rotect~on System (see Section 3.8) and the condensate transfer system.

1 , Simplified drawings of the ICS are shown in Figures 3.2-1 and 3.2-2. The makeup from the condens:ite transfer systrm to the emergency condensers is shown in Figure 3.2-3. A summu!. of data on selected ICS components is presented in Table 3.2-1.

3 . 2 . 3 O o c t - h During normal operiition the isolation condensers are in standby, and are placed

in service automatically when rirrded to provide heat transfer to the environment. The system is staned by opening nic~tor operated valves iri the condensate return lines. These valves can also be rerliwely actua~ed from the conuol room.

Steam jrom the reactoris piped to the tube side of the isolation coridensers. There are twointernal s e t s o l condensing tube bundles in each isolatidn 'condenser. Emergency cooling is accomplished by condensing steam from the reactor in the tube side of the condenser, which initially heats the stored water in the tank which eventually boils and vents to atmosphere.

Condensate flows from the isolation condensers back to the reactor through two of the recirculation loops. Flow is maintained by the thermal siphon established by the difference in density between the condensare in the return line and the steam in the supply line.

Steam generated i n ttie ~ s o l n t ~ o n condenser shell is normally non-radioactive. and 1s vented directly to atmosphere Water storage in the isolation condenser IS suffic~ent for about one hour and forty minutes w~thout makeup. The fire protection system can provide makeup to the condenser

3 . 2 . 4 . . . . . -enserr can provide adequate RCS heat removal

following a transient. Makeup to the condenser from the fin protection system is required after one hour and fony minutes of operation.

Due to RCS losses and stuinkagea,ssociated with ICS opention, makeup to the KCS eventually is required. The Control R q Drive Hydraulic System (see Section 3.7) can provided RCS makeup at high pressures. t

Oyster Creek

A. Isolat~on Condensers (2) 1. Rated duty: 205 x 106 B t u h at 1000 psig and SM'F 2. Rated capacity: 100% 3. Second.q side water volume: 22.730 gallons

3 . 2 . 6

A. Control signals 1. Automatic

The DC-powered motor operated valves on the condensate return line are autom;~tically opened by persistent signals of either high reactor pressure or

CS 1s put into ope opening of

ives can be actuated by remote manual means

s are Class 1E AC or DC, loads that can be esel generator or station battery, as described in

Section 3.6.

STEAM FKXl

REACTOR VESSEL

CONDFNSITi TO RFACTOH

~ .

. .. .

STEAM FKXl

rT IC IOR VESSEL

CONDENSATE TO RFACTOIi RCCIRCLJLA1V)N

14 34 TC-14-36 LOOPE . ~ . .

MAKEUP FROLI . .

Figure 3.2-1. Oyster Creek Emergency Condenser System

CONDFNSYR VI NT TO AlMOSl~lf 111

4

n CONDENSATE

I

Figure 3.2-2. Oyster Creek Emergency Condenser System Showing Component Location

Figure 3.2-3. Oyster Creek Condensate Transfer System Makeup to Emergency Condensers

.- - Table 3.2-1. Oyster Creek lsofation Condenser Sy Data Summary

. . for Selected Components -. .. . .

Oyster Creek , .

, , ,,. . ' . '

3.3 E : M ~ R ~ ; E X C ~ / C O R ~ ~ ~ , O O L ~ N G I;, , , : I ' , ., ,, . , , , . , ! , , ! . s ' ! 8 ( , , SYSTEM ECCSI ::>:! : , t~ ,

, , , ". ,. 3 . . , (,,, f i ' . i \ i . .'< I!, . !! :: ' $,;'cCS PC k&rratih s&fiUbsystems that perform eniergency coolant

injection and rwirculation furictions to mrtin.tain reactor core coolant inventory and adequate decay heat remcival f.ollii'$ng a,LOCA, or loksof the normal heat removal path via the main

; \;;,,, ' ' ' :, I condenser. i it _: , .

3 .3 .2 tion (ECI) function is performed by the, following

two ECCS subsystems:

- lsolat~ori Condt. ,ICS. see Section 3.2)

,., ' . ,., int where the c0r.e ;spray s y s t h pumps m the suppression pool to the reactor

ate supply of cooling water for the Core nd as a last reson water is provided by the System (Section 3.4) ha?:&? ~ f i ~ h a n g e r s the ultimate heat source: Tk isolation

ystem are shown in Figure 3.3-1 and 3.3-2. interfaces between this syste d the location of the ADS valvesiue shown in Section 3.1. 4 surnrnar selected ECCS components is presented in Table

. . , , S is in standby. Following a LOCA the core the core spray spargers. The normal water

, , ,

two main pumps. two boostel pumps, two outside the drywell, a spray spxrger, and s. The water supply for the system is held in to a common header. The header also feeds from each main pump flows into one of two

headers (,o,ne header for Loop,, I and another for Loop 11). The header connects the discharge of the miin pumps.10 thv.suction of the booster pumps, with a bypass line

3 . 3 . 4 coolant injection iEC1) and

The ECI system

corresponding booster pump with a suction on

18 9/89 . ,

. . , . , , . . Oyster Creek

mall LOCA is the following:

, , . , , , : ' i:ff ' ' , : I d Any four of five ADS,,^^!!!& $jc. ,y,~, d e n t to e ressurize the RCS. l i is pdssible that the coolant inventory coh:,dl fbnc'tibri[fdr some small LOCAs can be satisfied by low-capacity high-pressure injecrion system d c h as the control rod drive hydraulic system (CRDHS, see Section 3.7) Use of the, CRDIil$,in this role is not considered an ECCS function.

The ECR suqc$ss,crit~na for LQCAs are related to the ECI success criteria above. The core spray{sysien~ is essentially operating in a recirculation mode when drawing water from the supp$ssion pool.

For transients, the s u c ~ e s criteria for reactor coolant inventory control involve the following: . t .

, . 1 . ; , , , ; ,, ;,, , ; , , : : I . $ .'.I ;.>;; : ,

,.v ' E,$tWr the ~~blhf id~"cbr l~~hstY$yg~~m (sm Section 3.2). - Small LOCA mirigating:,:systems

, ' , : , , 5 13 ,' 3 . 3 . 5

. , 450 f t head (195 psid) ndem with one booster pump

3. ,,T,yp': horizont;\l. cbnmfugd , , : . ,

1, : , 7 1 $,$ ! I . ~ o r e . ~ ~ r ; t ~ ~ o q s t e r Pymps (4)

I . :Rated flo9:;37@ g ' :p @ 250 ft. head (108 psid) 1 Rated CnpGqy: in tandem with one main pump 3 . Type: hori~nt~l,,c$(pfugal

,, ,.,:< ' , %;

C . ~ u ~ ~ r e i s i o n P O ~ I (TO%$) I. Nonr~;r! wafer volu;qe. 87,660 ff3 2. Minir;num water vo;lupe: 82,500 ft3 , 3. Maxim;ni nonnal $mating temperature: 90 to 95OF (est) 4 . Design terpperirtun?;; 281°F 5 . Design,prgs$ure: , , 56 psig

, . , . ! 4.). D. Condensate storage Ti&

l . Capacity: . 525,000tggllons . , , i !

'I ; 3 . 3 . 6 'v , . , I

, . , , . ' 4, ,.,:I,,, .,: ' 1

, . ,,,, ,+,,:,,I.

started upon a high containment pressure or

pon a low leactor water level, coincident with high containm$ji presske. In addition, the core spray pumps must be

, , in operation. i ' 6 . Once reactor plr,dssure drops below 285 psig and the core spray pumps : arc operating,'the,four parallel normally closed isolation valves ( w o in

each loop) are opened. !, . . ,.

Oyster Creek

s can be actuated by remote manual means from the

d motor-operated valves are Class 1E AC from the emergency diesel generator or

erated valves. These valves do

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CFS 1D

Figure 3.3-1. Oyster Creek Core Spray System

. _ , . , . . . . , , , i -..- . , . . : . . . . . . . . . , . . . . ~ a . , , , , : . . . .

-- I RC 1 -

Figure 3.3-2. Oyster Creek Core Spray System Showing Component Location

Table 3.3-1. Oyster Creek Emergency Core Cooling System Data Summary for Selected Components

.- ~ . ...

I > .'A

. . . : - . .

I Oyster Creek

3.4.1 hn w j u r y t i o n with ~hcEmcrgency Service Water System complete

the heat trarisfer pgth from I tfi.e containmltit,to'he ultimate heat sink. :.TheCNS, is designed to reduce contaidment ttmperature and pre'sshre by deliverinn water' from the suppression -. pool, through heat exchangkrs, to spray headers inihe drywelTand toms.

3 .4 .2 . . . -f two redundant loom. each of which contain two 100% . . -

pumps in parallel, two 100% heat exchangers in pkdle l and independent sets of spray headers. Both loops share a common suction header from the suppression pool. Simplified drawings of the containment spray system are shown in Figures 3.4-1 and 3.4- 2. A summary of data on selected CNS components is presented in Table 3.4-1.

in standby. Follo eader, through heat exchangers and then

e ESW system removes heat from the

3.4.4 Successful containment heat removal can be achieved by either CNS train. For

a given train, one of two pumps must operate, taking suction on the suppression pool, and through one of two heat exchangers to the spray headers in the drywell

. . .

. , . t 3 . . , . , i

? I!#& !,:,,\:I. ,:.;\,12,

Figure 3.4-1. Oyster Creek Containment Spray System

Figure 3.4-2. Oyster Creek Containment Spray System Showing Component Location

Table 3.4-1. Oyster Creek Containment Spray S y for Selected Components

-. . ~ .~--. .

stcr Crerk

I 3 . 5 ONTKOL ( I & . . , ,. , 1 ._: . 3 . 5 . 1 n ,. , . , ~ , . . ,

" ' I % % k % % k % n and contrbl systems consist of the Reactor Protection System (RPS) and systems for the displby of piant information to the operators. The RPS ,monitors the reactor plaht. and alerts the operator to tAe corrective action before specified limits are exceeded. The RPS will initiate an automatic reactor trip (scram) to rapidly shutdown the reactor when plant conditions exceed one or more specified limits. It will

'also automatically actuate selected safety. systems based on the specific limits or 'combinations of (imits that are exceeded.

I

3 .5 .2 7 . . . The RPS includes sensor and transmitter units, logic units, and output trip

relays that interface with the conrral c,ircuits for components in the Contro1,Rod Drive .I:;,.H$~~,~aul.ic,:S~st,~.m. (See Sc,v&iOri ,3,6) V,.o,q~r ,Certain c i r c u m s j a n c e s ' . ~ o m p ~ s in other

safety'syst.~msyill ,gl.s.o be '?ctua!$d. instrumentation display systems consist of display pan~ l s that z q powered, by 12 , , .,,. wcr (see Section 3.6).

: , 1 : : ' , , , , . . , . 5 ,

' , ., , I. . . . r # , 3.5.3; , , . " ,* p, , , " !.>

I '

Oyster Creek C. Remote Shutdown

No information was found in the FSAR regarding a remote shutdown capability. Such a capability is expected to exist.

. . A. RPS

The RPS uses hindrance logic (normal = 1, aip = 0) in both the input and ourput logic. Therefore, a channel will be in a m p state when input signals are lost, when control power is lost, or when the channel is temporarily removed from service for testing or maintenance (i.e. the channel hss a fail-safe failure mode). A reactor scram will occur upon loss of control power to the RPS. A reactor scram is implemented by the scram pilot valves in the control rod drive hydraulic system (see Section 3.7). In addition to the scram function, RPS

1 control circuiuy; or operating the RPS or other actuation

ay send qualified persons into rn some other' remote control ofor control cenier). To make rs must .be,'8;ai1qble to the

, . . , , , , 2. .,I:! .:: , , . 1 : I ,.

120 VAC buses (see Seaipn 3.5). . , ; ,.

lays are powered from 120 VAC buses (see , , . ,

3 . 6 ELECTRIC POWER 1

, ,

, , , , , , 8 ..

SYSTEM . 4 I ' ;

..Oyster Creek

: ' , , ' . , . . . i

3 ! 4 ? 1 , ; .: ? , 8 , , ,', .., , . ~ s y s t e m j u b p l i . e s power to various equipmentyd systems

~ { ~ d c d ' f ~ r , . $ : ~ q n a j operation and/orre:sp& to accidents. The bniite Class 1E elecmc power system suppons fie operation of safety class systems and instrumentation needed to establish and maintain a safe shutdown plant condition following an accident, when the normal elecmc power sources are not available.

3.6.2 The onsite Class IE elecaic power system consists of two 4160 VAC trains,

two 460 VAC subsystems, two 125 VDC subsystems, and two 120 VAC subsystems. Two emergency diesel generators supply power to the 4160 VAC system for motor operated valve and pump operation. Three station baneries supply power to the 125 VDC system for normal switchgear control, turbine control, and various emergency functions. The 120 VAC system supplies power to the reactor instrumentation and pbtection circuits.

,, L ., . ,,,,.. '%".: :Simplified he - l ihe diagrams' of the electric'power syst'enis a?e shown in Figures 3.6-1.;,3.6-2, and 3.6-3. A summary of data on selected electric power system components is presented in Table 3.6-1. A panial listing of electrical sources and loads is presented in Table 3.6-2.

. , , . , , ,

3 . 6 . 3 ' J , , . t . =systemis n o d source of st&, sew& power under

both normal operatiqg and shutdown conc$tions. Auxiliary power is obtained from the main generator, through the station serviceuansformer connected to 4160 volt buses 1A and 1B. , , , , . ., , .

Upon loss of auxiliary power th; emergency diesel generatorsilreautomatically s t a n d , supplying power to 4160 VAC buses 1C and ID, which in turn supply power to 460 VAC buses l A l , 1A2, 1A3, IB!, 182, and 1B3 through a series of aansfomers. These buses constitute the two separate 460 VAC mains, distributing power to rest of the I ystem. , ! , , ,

The norm; source of power f~ r ' t he 125 VDC system are'thrc~ station batteries. Battery A can be energized by Banery charger B or MG Set A that are connected to motor

,, control center 1B2. Battery B can beenergized by Bancry charger B or MG Set B which is also connected to MCC 1B2. Battery C can be energized by either battery charger C1 or battery charger C2 which are both connqcted to MCClA2.

:,Thenarc eight 120 VAC,insaufient and protection bus& 'Three ofthese buses a:e supplied by ,motor generator sets connected to buses 1A2 and 182. Four more are supplied by ihe, s h e buses through qansformers. The last one issupplied by the DC system through a.fotary inverter.

. . ' . Pcdundant safety equipment such as motor driven pumps and motor operated valves arc supplied by different 480 .ybc buses. For the purpose of discussion, this equipment has been grouped into "load groups". Load group A contains components receiving er from bus IC. Load group B contains components powered by bus 1 D. . .

3.6.4 Basic system success criteria for mitigating nansicnts and loss-of-coolant

accidents are defmed by front-line system, which then m a r e demands on support systems. Electric power system success criteria are defined as follows, without raking credit for cross-ties that may exist bcnvccn independent load groups:

, Oyster Creek ..: 1 ,, ..,) , ' *: , . *. . . , , , ., . : ' , : : ! , ~ , - Ench C!ass'lE DC load group is sqppli* initially from its respectivtbiglery or

l i~m, jhe 'ba t t e ry charger aftrr the diesel ynernror is stirtet ' (the dicsel .. gcncr,ap? h.ave.sep?mtc s::L'::T.$ bat:qies:. ' E'aeCJass'1,E A C logd:'group IS !$$at4 from the non-Class 1E system and is

&pp!i$'b,oh its.qspective emerdericy+wer source (i.e. diesel generator) - ~bwcr~ism6utiot1 paths to essential loads are intact - Power to the battery chargers is restored kfore the batteries nn exhausted

A. Standby diesel generators (2) 1. Power rating: 2500 kW 2. Rated voltage: 4160 VAC 3 . Manufacturer: General Motors

' , , , / . , ,,,,,..,: !3,;1 ~tiqion,,b:a:ffe$,cs (3). . '?' ': . ' ,.a(> . .,,.,,,, ,.< ,.

1. Type:. had-acid: .; , . 2. Cells:,':6Q' ' ' ' . : . 3. Capadty; , . . . 90 minures supponingessentisl DC loads (estimated) . >,,, : , I : '

3 , ' ., i , !

matically started b . auxiliary power. 2. Rem~te manual

The, diesel genentors 'can be started, and many distribution canbe operated from the main ~ o n u o l room.

, . , . , . . :.

B.; ~ i e s e l Generator Auxiliary Systems , ,' The following auxiliaries are provided for the emergency diesel - ' Cooling . . , ,

Each diesel generator is providex$with its own closed cooli that,transfers heat to the ultimate heat sink by means

. , , radiator. Suppon from an e x t e m ~ s e ~ v i c c water system is n o t r q u i d - Fue!b~g , .

The, diesel fuel oil supply in the diesel fuel tank room is capable of suppoiting long-term diesel operation (approximately thrce-day supply). - Lubrication

ator has an independent lubricating oil system. , ,

: . .

ratoi has its 64; 125 volt starting. battery located in the 1 genentors are not dependent on the 125

, ,

., i . : . , , ,

the diesel room ventilation system have not I . :

. , . ,,

, .

31 ' 9/89 . .

, ,

. P O m R PUlEL V4CP.I. WNTlHXlS MSTRUYENI PANEL NO. 3. AND

~ i ~ u r e ' 3.6-1. Oyster Creek 4160 and 460 VAC Electric . ,

Power Distribution System . .

. . . . 32 9/89

, , , ,

. ,

FROM BUS 1A2

t

I 1 , 460 VAC MCC l A 2 1

I I 4 M VAC MCC 102 I

CONTAOC POWER TO4ldOV BUSES I&. 18. LND I D SWGR. DG4, SWITCHGEAR.

4 P V BUSES 181. 182. M D IN3 SWGR

Figure 3.6-2. Oyster Creek DC Power Distribution System I

3 3 9/89

dorm YEPATO UOI

.... ...... ~

ATE VACP.? ->-- .. , . .,

PANEL VACP.1 ',:

125 vCC PANEL 0

I

Figure 3.6-3. Oyster Creek 120 VAC lnstrumentatlon Power Dlstributlon System

34 9/89

Table 3.6-1. Oyster Creek Electric Power System Data Summary for Selected Components . .

-6-1. Oyster Creek Electric Power System Data Summary for Selected Components (continued)

ster..Creek Electric P r selected 'Componen

::. -Table 3.&>,!::g.ys.ter . . Creek ;Electiic:Power-.Sys.bm Data Summary ;: .: -fai Sel&ted.Lc&q$onfjnts . .

~. ~ . . . . . - . . . ..POWERSO(~ACE 1 . LOCATION

UNKNOWN

EMERG. LOAD GRP. AIB

: :;. Table 3.6-2. Partial Llsting ;of Electrical Sources ,and Loads . . > , .: at Oyster Creek , . . .. . . ,, . . . . , . . ! , : < \ . , , . , , ~ . . , ,

.., , . , . , . , . . i I able 3.6-2. Parlial L Electrical Sources ang Foads

k (continued) .. .. . . , , . ,

Table 3.6-2. Partial Listlng.:p!Electrical Sources a at Oyster , Creek . , . . , , . (continued)

. .

:,

able 3.6-2. Partlal Lls1l~g;of Electrical Sources at Oyster . . . . Creek . . , (continued)

:: ,

Table 3.6-2. Partial Llsling of Electrical Sources and Loads at Oysler Creek (continued)

COMPCIJENT LOCATION

4BOVSWGRM

COMP N P E

BUS

LOAD COhaPONENT ID

BUS 182

LOAD SYSTEM

EP

POWER SOURCE LOCATION

48OVSWGRM

EMERG LOAD GRP

0

POWER SOURCE

TR.lDlB2

VOLTAGE

460

Oyster Creek , .

3.7 TROL ROD DRI RAULIC SYSTEM! l i ,: i ; c ~ b p ) , . ,,. ' . ....

, . I , ( , , , .,..,, . , . ,, . 3 . 7 . 1 , . hl ,:, ,

-upplies pressutiied . , . , . . , water to operate and cool the control rod drive mccha+sms during normal operanon. .This system implements"a scram command from the reactor pmtecrion system (RPS), and drives control rods r The CRDHS flso can provide makeup wafer'to the RCS.

:.::,> ' ': , < " :

3.7.2 The CRDHS consists of two high-head, low-flow pumps, piping, filters,

control valves, one hvdraulic control unit f6r each control rod-drive mechanism, and inshumentation. at& is supplied from condensate and from the condensate storage tank. The CRDHS also mcludes scram valves, scram accumulators, and a scram discharge volume (dump tank).

A simplified drawings of the CRDHS is shown in Figure 3.7-1 and 3.7-2. Details of the scram ponion of a typical BWR system are shown in Figure 3.7-3 (adapted

;:f[om.R$. ,.,,. , l).::A:summary of data on sel&d,CRDHS components, i$presented in Table 3:7-1. , , . . , . . .

, , , , , . : ,

3.7.3 -tion the CRDHS pumps provide a constant flow for drive

mechanism cooling and system pressure stabilizanon. Excess water not used for cooling is discharged to the RCS. Control rods are driven in or out by the coordinated operation of the direction control valves. Insertion spccd is controlled by flow through the insen speed control valve. Rod motion may be either stepped or continuous.

A reactor scram is implemented by pneumatic scram valves in the CRDHS. An inlet scram valve opens to align the insert side of each control rod drive mechanism (CRDM) to the scram accumulator. An outlet scram valve opens to vent the opposite side of each CRDM to the dump unk (scram discharge volume). This coordinated action results in rapid insenion of control rods into the reactor. As shown in Figure 3.7-3, the scram ponion of the CRDHS bypasses the normal rod control portion of the system.

Although not intended as a makeup system, the CRDHS can provide a source of cooling water to the RCS during vessel isolation. It is noted in NUREG-0626 (Ref. 2). that this function is panicularly important for some BWRIl plants and BWRf2 (i.e., Oyster Creek) plants for which the CRDHS is the primary source of makeup on vessel isolation. The maximum RCS makeup rate of the CRDHS is about 200 gpm with both pumps operating (Ref. 3).

3.7.4 For the scram function to be accomplished. the following actions must occur in

the CRDHS:

- A scram signal must be transmitted by the RPS to the actuated devices (i.e., pilot valves) in the CRDHS. The pneumatic inlet scram valve and outlet scram valve must open in the hybaulic control units (HCUs) for the individual control rod (Irives. This is accomplished by venting the instrument air supply to each valve as follows: - Both m pilot valves in each HCU must be decnagizcd, or - Either backup scram pilot valve must be energized. A high- ressurc water source must be available from the scram accumulator in each H &J . . A hydraulic vent path to the scram discharge volume must be available and sufficient collection volume must exist in the scram discharge volume.

44 9/89

, . , , ! ; . ; . '

.,., , . . , , ?:Oyster Creek

A specified number of control rods must responds and insert into the reactor core (specific number needed is not known).

During isolation condenser operation, RCS makeup can be provided by one control rod drive pumps. I t may be possible to utilize the CRDHS to mitigate small LOCAs. In this case both pumps probably would be required. . -

A. Control rod drive pumps (2) 1. Rated capacity: 100% (for control rod drive funcrion) 2. Flow rate: about 120 gpm @ 1500 psig 3. Type: centrifugal

B. Condensate Storage Tank I . Capacity: 525,000 gallons.

and withdraw rods, or to inject water into the RCS

B . Motive Power The control rod drive pumps are Class 1E AC loads that can be supplied from the emergency diesel generator as described in Section 3.6.

3 . 7 . 7 3.7 Rercrences : .

1. N E D O - ~ ~ ~ O ~ A , "Additional. Information Required for NRC Staff Generic Report on Boiling Water React,ors," General Electric Comppny, December .1?,80. .> : , , , , . ' , . , . # , ! , ,

-0626, "Generic ~ v a l ~ a t i o n of Feedwater Transients and Small Break Coolant-AccidentsinGE-designed Operating Plapts b-4 Near-tern

ng License Applications," USNRC, January 1980. ' :' : '? ' ' " , , , ' \

ngton, R.M.. and Ott, L.J., "The Effect of Small-Capacity, High-Pressure Injection Systems on TQUV Sequences at Browns Ferry Unit One," WREGICR-3 . . . 179, Oak Ridge National Laboratory, Sept ' , , ,

,, . , , .

. ,

. ..

flgure 3.7-1. Oyster: Creek Control Hydraulic system : :

. ~ . . od Drive Hydraulic S

B I D R P S i l l Y R O T V U K S 1 4

W4E1 Y R L Y Y N K

4

I

YCm w:

VENT ++ $I i M R C T D U L CCNlm VNVES S C W mMIOtE K X W E ,I

E I M M T WATER T O R S ITH'S PATI( YAY BE S U O M O H SOYE F'LLWS)

S C R U I DlSCUIRGE W A E a F R D I OTHER IM

Figure 3.7-3. Simplified Diagram Of Portions Of The Control Rod Drive Hydraulic System That Are Related To The Scram Function

C m f f i W A I E I

W r W l j W I T E R

-. . . :..

Table 3.7-1. Oyster Creek Control Rod Drhe~~ydraul ic System .. ~.

. . . ~ ~. . Data Summary for Selected . -Components . .. . .

. .

. ~. ~.

COMPONENT fD

CRHSBA

CRHS-88

CST

COMP. TYPE

MDP

MDP

TK

POWER SOURCE

BUS 1 A 2

BUS 182

POWER SOURCE LOCATION

480VSWGRM

480VSWGRM

LOCATION

CRDRM

CRDRM

CST

. . ~ ~ . .

VOLTAGE'

460 -

460

EMERG. LOAD GRP. A

B

-

., . .

Oyster Creek , , ' . , .: . , > , , ' . ,. .

3 . 8 FIRE PROTECTION (FP) SYSTEM . , . (.. ; !

3 .8 .1 , . . . . system provides water to various plant locations for fire

fighting. In addition, the FP system is us4 ,a s a source of makeup water for , . the shell . sides : , \ . , , of the emergency condehsers. , ,,, . . . , .

3.8 .2 The fire protection system consis!$;of two diesel-driven pumps and the various

valves and piping required to deliver water to the plant. The source of water is a pond containing 7 .2 million gallons. Simplified drawings of the f i e protection system are shown in Figure 3.8-1 and 3.8-2. A summary of data on selected f i e protection system components is presented in Table 3.8- 1.

3.8 .3 -ration the FP system is in standby. kept under constant

pr,ess,u

Oyster Crcek 3. Lubrication

I t is assumed that each diesel-driven FP pump has an independent lubrication system.

4 . Starting Each diesel driven pump has its own 24 volt starting battery. These pumps are not dependent on the 125 VDC station baneries for starting.

yster ~ r & k F tern

> - 5 Figure;3.8-2. Oyster creek-Fire ~rotectibn System showing Component Location . .~

Table 3.8-1. Oyster Creek Fire Protection Syste ta Summary . .. . - for Selected Components

. .

-. . ~ . . -.

In P

. ! ' . 6; . . . . Oyster Creek

. , ,

. % , .. . . , .

3 .9 EMERGENCY SERVICE WATER SYSTEM (ESWS)

3.9.1 The Emergency Service Water System (ESWS) provide cooling water to the

heat exchangers for the ~ontainmen't 's~ray system. In conjunction with the containment spray system, the ESWS completes the heat transfer path from the containment to the ultimate heat sink.

3.9.2 The ESWS system consists of two redundant loops, each consisting of two full

capacity pumps connectid to a header that supplies two contahment spray heacexchangers. The source of water is the n w service water intake structure and water discharge is to the normal service water discharge. Simplified drawings of the Emergency Service Water Svstem are shown in Figures 3.9-1 and 3.9-2. A summarv of data on selected ESWS

-: c&nponents is presented $Table 3.9-1. .

,,.b,,t,, 5 < !. ,2,,, .. . . . .. 8, $,.,, .... .

3.9 .3 0oer;llian During normal operation the ESWS is in standby. When actuated. by low

reactor water level and high drywell pressure, the ESWS pumps take suction on the raw senice water, delivering water to the ESW heat exchangers and discharging it to the service water discharge line.

3.9 .4 The FSAR (Ref. 1) states that each ESW pump is 100% capacity. It is assumed

this means that a single ESW pump can cool both containment spray heat exchangers in the same loop. i

A. ESWS pumps (4) 1. Rated flow: 3000 gpm @ 100 ft. head (43 mid) 2. Rated capaciiy: 10@% 3. Type: vertical cenuifugal

B. ESW heat exchangers (4) I . Rated duty: 42 x 106 B t u h

A . Control Signals 1. Automadc

The ESWS pumps are s t a n d automatically on coincident low reactor water level and high drywell pressure signals.

2. Remote manual The ESWS pumps can be staned manually in the control room or locally at the pumps.

B . Motive Power The ESWS pumps arc Class IE A C loads that can be supplied from the emergency diesel generator as described in Section 3.6.

Oyster Creek

3.9.7

1. Oyster Creek FSAR, Section 6.2.2

Figure 3.9-1. Oyster Creek Emergency Service Water System

Figure 3.9-2. Oyster Creek Emergency Service Water System Showing Component Location

~.

9-1. oyster^ .Creek Emergency Service Wate stem Data Summary . . 'for Selected Components -.

8.. . : ., ,.,..

I . , , ,

",.I :

3.10 SHUTDOWN COOLING SYSTEM

Oyster Creek

3.10.1 :- . . . . . , , . The shutdown cooling system (SCS) provides for shutdown cooling of the

reactor:aftcr the RCS has been depressurized to less than 150 psig. The SCS transfers decay hcatto the ieacior closed co~bling'water system. This system.is not part of the Emergency Core Cooling System. ' : ! ! : ! ' '~ ' . . , . . : .

3.10.2 . . .

W y a single train system with three parallel SCS pumps and three parallel SCS heat exchangers. l l ,e system takes a suction on, and returns coolant to RCS recirculation loop E. Simplified drawings of the shutdown cooling system are shown in Figures 3.10-1 and 3.10-2.

3.10.3 -ation. the SCS is shutdown and isolated from the RCS.

Following reactor shutdown, RCS cooling is provided by steaming on the turbine bypass system and providing makeup to the RCS with the main condensate and feedwater system. The SCS is manuallv actuated durine RCS cooldown and devressurization when RCS oressure is less than'l50 osie and tcmoerature is less than 2 8 0 ' ~ . The SCS is desierned . - kith sufficient capacity to remove deciy heat being generated by the core 24 howlafter shutdown and hold RCS temperature at 125'F (Ref. 1).

3.10.4 . . Not determined.

A. SDC pumps (3) 1 . Type: Cenmfugal 2. Capacity: 3.000 gpm (approx.) at unknown head

B. SDC heat exchangers (3) , , 1. Type: Horizonlal u-tube 2. Heat removal capacity: 11 x 106 Btuhr

A. Conml Signals 1. The SCS is controlled from the reactor control room 2. An interlock prevents opening the isolation valves between the RCS and

SCS under any of the following conditions: - RCS pressure greater than 150 psig - Low-low RCS water level - High drywell pressure B. Motive Power

The SCS pumps can be powered from the diesel generators.

C. Others The reactor building closed cooling water (RBCCW) system provides cooling water to the SCS heat exchangers and to the SCS pump bearings.

Oyster Creek

3.10.7 Section 3.10 Relerences

1. Oyster Creek FSAR, Section 5.4.7.

I REClRUATlOW LOOP E I

CLOSED C o a INQ

WATER. SYSTEM

RETWIN TO AUCTOR CLOSED COOLING WATER SYSTEM

e . m w Figure 3.10-1. Oyster Creek Reactor Shutdown Cooling System

Figure 3.10-2. Oyster Creek Reactor Shutdown Cooling System Showing Component Locations

4. PLANT IhFORhlATION

4.1 SITE AND BUILDING SUhlMARY The Oyster Creek nuclear power plant is located on a site of approximately 800

acres near the Atlantic Ocean within the State of New Jersey. The site is in Ocean County. about 2 miles inland from the shore of Bamegat Bay. The site is approximately 9 miles south of Toms River, N.J., about 50 miles east of Ph~ladelphia. PA, and 60 miles south of Newark, N.J.. Figure 4-1 is a general view of the plant and vicinity (from Ref 1).

The major s m c ~ r c s at this unit include the reactor building, turbine building. office building, nwntcnance building, and materials warehouse. A site plot plan is shown in Figure 4-2,-

The primary containment consists of two large chambers, the drywell and the torus. The drywell houses the reactor vessel. the reactor coolant recirculating loops, and other components associated with the reactor system. I t is a 70 f t diameter spherical steel

; shell with,a 33 ft. by 23 ft. high cylipdficdl steel s k i ] cx:rndingfrm the top. The torus is . . % I a steel shell, located below and ar@un'$,&e base of the drywell. 1bhrs a-major diameter of

101 ft., a chamber diameter of 30 ff,and is filled to approximately 12 ft,. dlpth with water. The NO chambers are interconnected ttlrough ten vents, 6.5 ft. in diameter, e ually spaced

building is shown in Figure 4.3. . , 8 . .

8 around th~circumference of the torus.An elevation drawing of the Oyster reek reactor . , The turbine building is a reinforced concrete structure which houses the power

equipment, related auxiliijry systems, the plant control room and electrical ' , 8 !

FACILITY LAYOUT DRAWINGS Figures 4-4 through 4.1 1 are simplified building layout drawings for the Oyster

actor building. The turbine and office building layouts are shownin Figures 4-12 . The intake structure is shown in Figure 4-17. The freshwaster pumphouse is shpwn in Figure 4-18, and the diesel generator building is show^^ in Figure 4-19. Major rwms, stairways, elevators, and dooways are shown in the simplified layout drawings, howcver, many interior walls have been omitted for clarity. Labqls printed in uppercase correspond to the location codes lissed,in Table 4-1 and used in thccomponcnt data listings a.nd system c/nwings in Section 3. Spde addjtionul labels are inclGded for i,nformation and urc printed in lowercase type.

A listing of component^ b;location is presented infable 4-2. Com onents P i n h k d in Table 4.2 are those found in the system data tables in',Scction 3, then ore this table is only a partial and equipment that atcd in a pardcular

' . room or areaof the plant. :

SECTION 4 ' J.3, !,

1 . ticddleson, F.A., " ata and Safety Features of crcial Nuclear Power Plants," OR 4 5 , Volun~e 1, Oak Ridge al Laboratory. Nuclear Safety Info enter, December 1973.

, . . ,

Figure 4-1. General View of Oyster Creek Nuclear Power Plant and Vicinity.

CALLED WRlH

Flgure 4.2. Oyster Creek General Slte Plan

t CALLED NORTH

1

, .

c,, e4-4. Oyster Creek ~eactor Building - ~levatlon],(-), 19'-6"

,,. . I / , . . 1

t CALLED NORTH

Figure 4-6. Oyster Creek Reactor Bullding - Eievatlon 23'-6"

t CALLED NORTH

3

It Rulmsd A~rlock

CALLED NORTH

Figure 4-7. Oyster Creek Reactor Bulldlng - Elevatlons,33'-5" to 38'-0"

CALLED NORTH

Figure 4-8. Oyster Creek Reactor Building - Elevation 51'-3" 72 9/89

Spent Fuel /

PWI Heat Exchangers

Area

Reactw water Cleanup Area

, ,

Figure 4-9. Oyster Creek Reactor Building - Elevation 75'-3" 73 9/89

New Fuel: storage -

Standby Llquld ,

Contfol system

Emergency Condensers

t CALLED NORTH

Figure 4-10. Oyster Creek Reactor Building - Elevation 95'-3" 74 9/89

, ; i s ,. . 5..

Spent Storage

Fuel Pool

Reactor Rail Cawty

Steam Separalor

Slorage Cavlty rq ETl:J(

t CALLED NORTH

i

Figure 4-11. Oyster Creek Reactor Building - Elevation 119'-3"

LOEAYCOR

Condenser pq

Condenser p-l Feewater

and Condensate

Pump Area

U '

Closed Coolmg Water

System Area

Ejector Room

+ o CRD Pump Room eanor h ~ l d ~ n g I ElevaDon (.)1'.11'

Reactor Building

. .

I;i ~buipmentDram larik Room, 3ea$or Building i t Eyvation(.p'.B'

t CALLED NORTH

P TUNNEL

Nole. Diawlng s d e is dinerent man iw reactor hiking drawinpc i

~ & e 4.12. Oyster Creek ~"rb lne Building - Elevation q'-0" and 3'-6" i . , . .

i t ' ' !

Main Condenssr

Area

Main steam Line Area D

Track

Feedwater Heater

Slab at grade

t CALLED NORTH

Cwl~ng Bay , .

. , , ,

Noto Drawq scale 1s dlHerent man lor reactor building w i n g s

Figure 4-13. Oyster Creek Turbine and Office Buildings -'Eievatlons 23'-6" to 27'0"

Main Turbine

Nolo Drawlnp s a l e I& M e r g n l man for relccoc butldlng Qmlnps Figure 4-14. Oyster Creek Turblne end Office Bulldlngs - Elevation 35'-0"

- t CALLED NORTH

Main Turbine

To Reacmr Buildmg Elevauon 51'4'

Reactor Building

Note Draw~ng scale 1s dtlerenr man lor mIcW buikhng Qawmgs

Flgure 4-15. Oyster Creek Turblne and Offlce Bulldlngs - Elevallon 46'4"

Cable Tray Bridge , .

EL. About 80'

. ,

Nole Drawing scale is dtlerenl man lor maclor bullding draw~ngs

Figure 4-16. Oyster Creek Turblne and Office Bulldlngs - Roof Elevations

, . , .>

8 ,? . , * .

. . , t ' CALLED Sample NORTH Pool

. .

HB ;w Pump, Clrculabng

Weler pump6

C~rculaung Water Pumps

Figure 4-17. Oyster Creek lnlake Structure, Elevation 6'0"

I

~ i i & l d r l v m Firs Pump 2

0 Pumps 0

Fuel Tank

Figure 4-18. Oyster Creek Freshwater Pump House, Elevation 93'0" i( 1

. , : I .I . j i 3.. :

82 ''I 8 , , . , / .

9/89

Owel

Tank

I DFTKRM DGl RM

Diesel Generator

2

t CALLED NORTH

, ' ' ,. ,. .. . , > , : ,

Table 4-1. ~eflnlt lon' df.0yster Creek Building and Location . : , , Codes , , I

I , /

Motor Control Center #1A21A, located on the 23' elevation of the Reactor Building - northeast corner Motor Connol Center #1A21B, located on the 23' elevation of the Reactor Building - n'brtheast comer Motor Control Center #IAB2, located on the 23' elevation of the Reactor Building - south side Motor Control Center #IAB2A, located on the 23'elevation of

. , the Reactor Building - nonheast comer , >-.: ,IT3 .,., c*> ; . . . " . ,.,, v. , . Motor Control Center #1B2B, located on the 23'elevation of the Reactor Building

4160V Switchgear Room #1C, located on the 23'elevation of the Turbine Building

4160V Switchgear Room #ID, located on the 23' elevation of the Turbine Building

I 4160V Switchgear Room #lAB, located on the 23' of the Turbine Building

I I

23' elevation of the Reactor Building

51' elevation of the Reactor Building ,

75' elevation of h e Reactor Building 1

95' elevarion of the Reactor Building I

480V Switchgear Room, located on the 23' elevition of the Reactor Building - northwest c o m a Battery Room #AB, located on the 35' elevation of the Office Building

Battery Room #C, locatcd on the 23' elevation of the Turbine Building - southwest comer Condenser Bay, located on the 0' elevation of the Turbine Building

I

Control Room, locited on the 46' elevation of t i e Turbine Building - northwest comer

, , .,, : , , , , .>A$ , , . , ,

, , :. . . . . . , . , , , . . , ' ' ; Table 4-1. Definltion bf O y s t e r C r e e k Bui ld ing and

~. . Loca t ion Cpdes . L : > . . , ( C o n t i n u e d ) . . , :,.,

18. CRDRM Control Rod Drive Water Pump Room, located on the -1' elevation of the Reactor Building - northwest comer

19. CSR Cable Spreading Room, located on the 36' elevation of the Turbine Building - northwest comer

20. CST Condensate Storage Tank, located outside in Yard

nter #DCI , located o uthwcst c o m a

. .. .

ter #DC2, located o the Reactor Building

23. D m Diesel Generator Fuel Tank Room, located in the Emergency Diesel Gencrator Building

24. DGlRh4 Diesel ene era to; Room #I, located in the ~ m e t ~ e ' k ~ Diesel Generator Building

25. DG2Rh4 Diesel Generator Room #2, located in the Emergency Diesel Generator Building

26. HILOCR High/Low ConduCtivity Room, located on the 0' elevation of the Turbine Building -east side

27. INTAKE Emergency Service Water Intake Srmcturc, located next to the Intake Canal : '

28. LOBAY Lubrication Oil Bay, located on the 0' elevation of the Turbine Building - north side

29. LOBAYCOR Lubrication Oil Bay Conidor, located on the 3' elevation of the Turbine ~uilding '- 'cast side

30. NECNRh4 Northeast Comer Room, located on the -19' elevation of the Reactor Building

31. NWCNRh4 Northwest Comer Room, located on the -19' elevation of the Reactor Building

32. RC Reactor Containment, includes both drywell and torus

3 3 SDCHXRM Shuidown Cooling Heat Exchanger Room, the 51'3" elevation of the Reactor Building

9/89

T a b l e 4-1. Definltlon of O y s t e r Creek Building a n d Locat ion C o d e s (Cont inued)

34. SDCPMPRM Shutdown Cooling Pump Room, on the 38' elevation of the Reactor Building

35. SECNRM Southeast Comer Room, located on the -19' elevation of the Reactor Building

36. SWCNRM Southwest Comer Room, located on the -19' elevation of the Reactor Building

Pipe Tunnel, extends between the Turbine Building and Reactor Building - con,jAlHS'tlnergency service watcr,$ystem-piping , .

,.

, . , . , >I,.,,

Table 4-2. Partial Lls nents by Location

, . ;;$ , ! I ' .,

.. , ,, .

I

, , , , , ., I I ' I 1

, . , . , ! , i,'

. , , .'., , ! . ..

Table 4.2. Partial Llstlng o f :Components by Locatlon at Oyster Creek) (c' ontinued) .; ,'.,.,

i I:..,

, ,

Table 4-2. Parllal Llstlng of Components by Location at Oyster creek: (continued) . I . . . ~ . , ,

. . . . >.;

Table 4-2. Partlal Llstlng of Componenls by Locatlon at Oyster Creek (continued)

LOCATION

FC

W

I I 1 UNKNOWN ATSIT4

I .

RC l RCS I RCS.RV I RV

COMPONENT ID SYSTEM

h ?C

RC L

SECNRM

COMP PlPE

8 ,

RCS

RCS

RCS

XC

xc

CNS

RCS-17.54

RCS.17-19

MOV

MOV

R C S 1 6 1

EC.14.36

Ec.14.37

CNS21.3

MOV

MOV

MOV

M O V

Table 4-2. Partlal Listing of Components by Location at Oyster Creek (continued)

Oyster Creek

5. UlBLIO(;RAPIiY FOR OYSTER CREEK

1. "Final Environmcnul Statement Related to Operation of Oysler Creek Nucieiu Generating Station", USNRC Directorate of Lrcensing. December 1974.

2. NUREG-0822, "Integrated Plant Safety Assessment - Systematic Evaluation Program - Oyster Creek Nuclear Generating Station" USNRC, July 1988 (Supplement 1).

3. NUREGtCR-1015, "Analysis of Boring and Fouling Organisms in the Vicinity of the Oyster Creek Nuclear Generating Station", Lchigh University, 1979 (and eiulier repons with the same title: NUREGtCR-0896 and NUREGICR-0812).

4. NUREGICR-3446. "Ecological Studies of Wood-boring Bivalves in the Vicinity of the Oyster Creek Nuclear Generating Station". Academy of Natural

.,.., , .,.. , , . Sciences of Philadelpi-tia,. October 1983 (and .eqrlier progress report NUREGICR-2727). ,; ',' ',,,:': , ,

5 . PLG.OlO0. "Oyster creek Piobabilistic Safety ~ n a l ~ { i s " . Pickard. Lowe and Garrick. Inc.. August 1979 (draft).

,!.:

Oyster C r u k

APPENDIX A DEFINITION OF SYhiBOLS USED IN T H E SYSTEM.AND

L ~ Y o u r DRAWINGS , , j.j,

The s~mplified system drawings arc accurate representations of the major flow paths in a system and the imponant interfaces with other fluid systems. As a general rule, small fluid hnes that are not essential to the basic operation of the system are not shown in these drawings. Lines of this type include instrumentation lines, vent lines, drain lines, and other lincs that are less than l/3 the diameter of the connecting major flow path. There usually are two versions of each fluid system drawing; a simplified system drawing, and a comparable draw~ng showing component locations. The drawing conventions used in the fluid system draw~ng\ are the following:

,,,,,-,I; . . , . ,..., ; 'd,, . ,,.. . . ,~ . . , , ,. - Flow genemlly is left to right. - Water sources are located on the left and water "users" (i.e.. heat loads) or ~~~~ ~~ . .

disch:trge paths are located on the right. - One exception is the return flow path in closed loop systems which is right to left. - Anorher exception is the Reactor Coolant System (RCS) drawing which is "vessel-centered", with the primary loops on both sidqsof the vessel. - Horizontal lines always'dominate and break vertical l ines. ' , :

. : , >; , . . :,.. , , , . , . . ,. , I

Con~ponent symbols used in the fluid system drawing , . ) ) i ' A-I . . , . .; - Most valve and pump'symbols are designed distinguish among similar components based on their support system requirements (i.e.. electric power for a motor or solenoid, steam to drive a turbine, pneumatic or hydiaulic source for valve operation, etc.) - Valve symboli allow the reader to distinguish among valves that allow flow in either direction, check (non-return) valves, and valves that perform an overpressure protection function. No attempt has been made w 2cfine the specific type of valve (i.e,., as a globe, gate, butterfly, or other specific type of valve). ! :'., . , I ., - Pump symbols distinguish between cenmfugal andpositive displacement pumps and between types of pump drives (i t . , motor, turbine, or engine).

- Locations are identified in teims of plant location codes defirid in Section 4 of this Sourcebook. .) ;!: : - Location is indicated by shaded "zones" that are not i,ntended to represent

the actual room geometry. - Locations of discrete components represent the actual physical location of the component. - Piping locations between discrete components rqpresent the plant areas through which the, piping passes (i.e. including pipe tunnels and underground pipe runs). - Component locations that are not known arc indicated by placing the components in an unshaded (white) zone. - The primary flow path in the system is highlighted (i.e., bold white line) in the location version of the fluid system drawings.

'. 8 , .. , 94 9/89

OY ster Creek A 1.2 . . ,Electrical System Drawings! " 2 : -

,,, , I , I , ,'! ..,. ,. , ' . ; .,. . ,> . , . , I , . . , i ( .

.A,.i .(( The elictric power system drawings focus on the Clas pl:ini!s'rlcctiic power system. Se itrate dnwiiig'S;iire provided for the f ofihc Class IE system. There o ten are twd:Versions of each elecm simplified system drawing, and a compxablc b w i n g showing c o ,.

. , , ,., rbtivcntioris usEd in the electrical sjht~mdrawings arc the , . , . . ,.. ,.?>, ~:.:!': ?, - Flow generally is top to bottom - In the AC power drawings, th,e interface with the switchyard and/or offsite

grid is shown at the top of the drawing. - I n the DC power drawings, the batteries and the interface with the AC

power system are shown at the top of the drawing. - Vertical lines dominate and break horizontal lines. symbols used in the:elrctrical system drawin ed in

?,+!W.~ 4,. \, . , , .; t ' ,,

- Locations are identified in terms of plant location codes defined in Section 4 of this Sourcebook. - Locations are indicated by shaded "zones" that are not intended to represent

the ictu;~l room geometry; - Locations of discrete comp present the actual ph the component. , . . , - The electrical connections (i.e.',.cable runs) between disc as shown on the electrical systepdrawings, DO NOT represent the actual cable routing in the plant. , t : . . ' / I . . . , ,, , - Component locations that are not known are indicated by placing the discrete components in an unshaded (white) zone.

A2. SITE AND LAYOUT DRAWINGS # "

A2.1 Site Drawings . .

A general view of each reactor site and vicinity is presented along with a simplified site plan showing the arrangement of the major buildings, tanks, and other features of the site. The general view of the:reactor site is obtained from ORNL-NSIC-55 (Ref. 1). The site drawings are approximatelyto scale, but should not be used to estimate distances on the site. As-built scale drawings should be consulted for this purpose.

Labels printed in bold uppercase correspond to the location oodcsdefined in Section 4 and used in the component data sand system drawings additional labels are included for informati e printed in lowercase

A2.2 . Layout . Drawings ,. , .

Simplified building layout dra developed for the p that contain components and systems th bed in Section 3 Generally, the following buildings are included: reactor building, auxiliary building, fuel building, diesel building, and the intake structure or pumphouse. Layout drawings generally are not developed for other buildings.

Symbols used in the simplified layout drawings are defined in Figure A-3. Major rooms, stairways, elevators, and doorways tire shown in the simplified layout drawings however, many interior walls have , been . . , omitted for clarity. The building layout

Oyster Creek drawings, are approximately to scale, should not be used to estimate room size or distances. As-built scale drawings for should be consulted his purpose.

Labels printed in uppercase bolded also correspond to the location ccdes defined in Section 4 and used in the component data listings and system drawings in Section 3. Some additional labels are included for information and an printed in lowercase type. . . A 3 . APPENDIX A REFERENCES

1. Heddleson. F.A., "Design Data! and Safety Features of Commercial Nuclear Power Plants.", ORNL-NSIC-55, Volumes 1 to 4, Oak Ridge National Laboratory. Nuclear Safety Information Center, December 1973 (Vol.l), January 1972 (Vol. 2). April 1974 (Vol. 3). and March 1975 (Vol. 4)

MANUAL VALVE . XV ** (OPENKLOSED) MANUAL HOH.RETURL VALVE . XCV (OPEHICLOSED) YO1OO.OPERATED VALVE . MOV -+ YOTOR.OPERATED (OPEN,CLOSED) ¶.WAV VALVE . W V (CLOSED PORT YAY VARV)

/ I > I , , i "

r BOLENOID~OPERATED VALVE r 8OV IOLENOIO~OPERATEO ( O P I N I C L O I E D ) &WAV VALVE . SOV

(CLOSED PORT YAV VARY)

ncv (OPEMICLOSED)

PNEUMATIC VALVE . NV IC NON.RETURN (OPE N,CLOSEO)

POWER OPERATED RELIEF VALVE. F POWER.OPERATED RELIEF VALVE. SOLEHOID~PILOT TYPE . PORV PNEUMATICALLV OPERATED . PORV (CLOSED) OR DVAL.FUHCTION SAFETI IRELIEF VALVE SRV (CLOSED)

CEHTAIFUGAL CEWTRIFUOAL MOTOR~ORIVEN PUMP MOP TURBINE.OAIVEH PUMP - TOP

POSITIVE DISPLACEMENT YOTOR.DRIVEN PUMP . MDP

. . , ' . ' ' , , . ; d ; . '

Figure A-1. Key To Symbols In Fluid System Drawings

97 9/89

0 : , D;;, P W R I W R MAIN CONDENSE* . C O N 0 REACTOR VESSEL . RV ... , , . . . , , I , ? d : . > .- 9 1 , .

, $ 8

HEAT EXCHANOER. HX MECHANICAL DRAFT COOLINO TOWER

I 1TEAM.TO.WATER , . , , . , ,:. ' OR WATER.TO.STEAM HEAT EXCHAHOLR (I.E. FEEOWATER ."..

HEATER, DRAIN COOLER. ETE,) . - H I

,Alq,COOLINO UNIT . ACU

, ,

43- FILTER . FLT

Figure A-1. Key To Symbols In Fluid System Drawings (Continued)

A.C. DIESEL OENERATOR . W OR A.C. TURBINE GENERATOR . 10

CIRCUI1.BREAKER . CB (OPENICLOSED) I...{ 0. /--.I INTERLOCKED CIRCUIT BREAKERS . CB

5 . , , , , . , ' i i ' ) , , , .:.

, , SWITCH . SW

AUTOMATIC TRANSFER SWITCH . I T S DISCONNECT DEVICE OR (OPENICLOSED) MANUAL TRANSFER SWITCH . MTS

..~,. . .

SWITCMOEAR BUS . BUS (BUS NAME) CONTROL CENTER . MCC

DlSTRlEUl lON PANEL . PNL

' + ' BATTERY CHAROER (RECTIFIER) . OC INVEPTER . INV

ELECTRIC MOTOR . MTR 8 YOTOR OENERATOR - UG

Figure A-2. Key To Symbols In Electrical System Drawings

STAIRS U . up 0 s Down

@ SPIRAL STAIRCASE

, .

LADDER . .

,,..,..,, , ,,,. t ...,.. .. . . . D s Down

HATCH OR GRATING DECK

OPEN AREA (NO FLOOR)

--o- PERSONNEL DOOR I-IEQUIPMENT DOOR

t FENCE SINE

Figure A-3. Key To Symbols In Facility Layout Drawings

, ,, . ,. , .. . . , , . , ,

, , , ,, , . , . . , : .: , ..,. Oyster Creek

. . APPENDIX B

.:b,.

. . ,.. - . ,

DEFINITION O F TERMS USED IN T H E DATA TABLES .,. ..

Terms appearing in the datiiiablis .!? ,,,..,. %,. i n Sections 3 and 4 cbook are fol!ows: . . .,,:

i:', .::, ',,,.

SYSTEM (also LOADSYS'EM) - ~l\ ,?o&ponents associated with a particular system description in the Sourcebook have the Same system code in the data bdse. System codes used in this Sourcebook are the following:

QXk rA5mb.n RCS Reactor Coolant System XC Emergency (Isolation) Condenser System ECCS Cooling Systems (co

ch's EP Electric Pp,wer System CRHS Control Rod'Diive Hydraulic System FS Fire Protection System ESW Emergency Senice Water System

COMPONENT ID (also LOAD COMPONENT ID) - The component identification (ID) code in a data table matches the component ID that appears in the corresponding system drawing, The component ID generally begins with a system preface followed by a . component number. The system preface is not necessarily the samc as the system code

:. -&scribed above. For component IDS, the system preface corresponds to what the plant :;.,i.l$tlls@e. component (e.g. HPI, RHR).,,&/xample is HPI-730, den,ptlgg.,y,@e number 'i't730 in the high pressure injection systenil: which is pan of the ~ ~ ~ ~ : ; ~ j ~ t k c o m ~ o n e n t

number is a contraction of the component number appearing in the plant piping and instrumentation drawings (P&IDs) and elecmcal one-line system drawings.

LOCATION (also COMPONENT LOCATION and POWER SOURCE LOCATION) - Refer to the location codes defined in Section 4.

COMPONENT TYPE (COMP TYPE) - Refer 10 Table B-1 for a list of component type codes. . ,

POWER SOURCE - The component ID of the power source is listed in this field (see COMPONENT ID, above). In this data base. 3 "power source" for a particular component (i.e. a load or a distribution component) is the next higher electr&al distribution or generating component in a dismbuiion system. A single component may have more than one power s o w (i.e. a DC bus powered from a battery and a battery charger).

POWER SOURCE VOLTAGE (also VOLTAGE) - The voltage "seen" by a load of a power source is entered in this field. The downstream (output) voltage of a aansfomer. inverter, or battery charger is used.

EMERGENCY LOAD GROUP (EMERG MAD GROUP) - AC and DC load groups (or clecmcal divisions) are defined as appropriate to the plant. Generally, AC load groups are identified as ACIA, ACIB, etc. The emergency load group for a third-of-a-kind load (ix. a "swing" load) that can be powered from either of two AC load groups would be identified as ACIAB. DC load group follows similar naming convendons. .

, . , , . I ,,

,. .. . , . , , . . , , ...,,'

TABLE B-I. CO\IPOSEST TYPE CODES , < ' , . . . . ,%. .

i\ . .. ',

VALVLS: - Motor-operated valve MOV Pneumatic (air-operated) valve NV or AOV Hydraul~c valve HV Solenoid.operated valve SOV hlanual valve XV Check valve CV Pneumatic non-retum valve NCV Hydraulic non-return valve HCV ~r;fety valve Dual function safery[F3i~tvalve Power-ooerated reliefwhve (pneumritic or solenoid-operated)

PUMPS: Motor-driven pump (cenmfugal or PD) Turbine-driven pump (pnmfugal of PD) Diesel-driven pump (cegmfugal of PD)

OTHER FLUID SYSTEM C,O.MPOSEhTS: Reactor vessel Steam generator (U-tube or once-through) Heat exchanger (w water HX, or water-to-air HX Cooling tower Tank Sump Rupture disk Orifice Filter o r strainer Spray nozzle Heaters (i.e. press

.. , . , !

V E h ' L A T l O N S Y S T E ~ ~ C ~ ~ P O ~ T ~ T S : Fan (motor-driven, , y y b ~ e ) Air cooling unit (air-rewater M, usually including a fan) ,. , i ,. . Condensing (air-

EMERGENCY POWE Diesel generator Gas turbine gene Bancry

MDP TDP ' DRP

TANK or TK SUMP RD ORIF E T - - - S N I HTR

FAN ACU or FCU

TABLE B-1. COMPONENT TYPE CODES (Continued)

c O \ f P m 1 m M P TYf"

ELECTRICPOWER DISTRIBUTION EQUIPMEhT: Bus or switchgear BUS Motor control center MCC Dismbution panel or cabinel PNL or CAB Transformer TRAN or XFMR Battery charger (rectifier) BC or RECT lnvener INV Unintermptible power supply (a unit that may UPS include battery, battery charger, and invener) Motor generato Circuit breaker

, ./ , , . .,. Switch Automatic trans Manual transfer switch

Han an interesting 2.206 hearing on Oyster Creek today. I had not focused on thislanguage in the response until today:

 During a PMH flood, the circulating water and service water pumps will becomeinoperable and, thus, emergency plant procedures have been instituted which requirethe plant to be shutdown when flood waters reach a pre-determined level, as to ensurethe capability for safe shutdown under either normal or abnormal conditions.

 Is that normal or is this unusual? Richard   

From: Lamb, John [mailto:John.Lamb@nrc.gov] Sent: Monday, November 26, 2012 4:20 PMTo: Richard WebsterCc: Khanna, Meena; Burnell, Scott; Banic, Merrilee; Bergman, Thomas; Evans, Michele; Mensah, Tanya;Hair, Christopher; Sheehan, Neil; Screnci, Diane; Pelton, David; Kulp, Jeffrey; Hunegs, Gordon; Dacus,Eugene; Weil, Jenny; Lund, LouiseSubject: Oyster Creek 2.206 Petition - PRB Immediate Action Decision Importance: High  Dear Mr. Richard Webster:I have been assigned as the Petition Manager for the 10 CFR 2.206 petition you submittedto the NRC on November 19, 2012, regarding your concerns on Oyster Creek.Section 2.206 of Title 10 of the Code of Federal Regulations describes the petition process– the primary mechanism for the public to request enforcement action by the NRC in apublic process. This process permits anyone to petition NRC to take enforcement-typeaction related to NRC licensees or licensed activities. Depending on the results of itsevaluation, NRC could modify, suspend or revoke an NRC-issued license or take anyother appropriate enforcement action to resolve a problem. The NRC staff’s guidance forthe disposition of 2.206 petition requests is in Management Directive 8.11, which is publiclyavailable. I have attached it for your reference.In accordance with Management Directive 8.11, Part III.A.1 (page 7), the Petition ReviewBoard (PRB) met internally on November 26, 2012, to discuss the request for immediateaction regarding your emergency 2.206 petition regarding lack of adequate protection ofsafety at Oyster Creek. The PRB denied the request for immediate action to take emergency enforcement actionto prevent Oyster Creek from starting up from its refueling outage, because there was noimmediate safety concern to Oyster Creek, or to the health and safety of the public for thefollowing reasons:

(1) On November 13, 2012 (Agencywide Documents Access Management andDocuments System Accession No. ML12319A627), the Federal EmergencyManagement Agency (FEMA) concluded that offsite radiological emergencypreparedness is adequate to provide “Reasonable Assurance” and that appropriatemeasures can be taken to protect the health and safety of the public, in the event of

mailto:John.Lamb@nrc.gov

a radiological emergency at Oyster Creek in Ocean County, New Jersey.(2) Currently, 3 emergency notification sirens are inoperable out of a total of 42emergency notification sirens, which does not exceed Exelon's reporting thresholdof 25 percent or more sirens out of service. Exelon is working to restore the 3inoperable sirens. FEMA’s assessment determined that, in the areas where thesirens were determined to be inoperable, the FEMA-approved backup notificationmethod of route alerting could be conducted, if needed.(3) Hurricane Sandy did not exceed Oyster Creek’s maximum flood level due toprobable maximum hurricane (PMH). As reported in the Oyster Creek Final SafetyAnalysis Report (FSAR), Subsection 2.4.5, the maximum flood level due to PMHwill be at elevation 22 feet (ft) mean sea level (MSL). The plant grade, elevation 23ft MSL, is one foot above the PMH flood level. Therefore, the flood will not find itsway into the plant buildings, the floor levels of which are generally 6 inches abovegrade at elevation 23 ft and 6 inches. The circulating water intake structure, with itsdeck at elevation 6 ft, will be under water. This deck supports, apart from the otherequipment, the circulating water pumps and the emergency service water pumps. During a PMH flood, the circulating water and service water pumps will becomeinoperable and, thus, emergency plant procedures have been instituted whichrequire the plant to be shutdown when flood waters reach a pre-determined level,as to ensure the capability for safe shutdown under either normal or abnormalconditions.(4) During a planned, routine inspection program, Exelon discovered control roddrive return nozzle safe end to pipe weld indications. These indications weredetermined to be surface in nature and did not result in any leakage. Exeloncompleted a structural weld overlay in accordance with the ASME Code.

Per Management Directive 8.11, all of the information in your petition will be made public,including your identity. I would appreciate if you could advise me by November 30, 2012,if you:

· Agree to the NRC’s processing your request under the 2.206 process. · Request an opportunity to address the PRB. If you would like to meet in person, I

will need to schedule a public meeting at the NRC Headquarters in Rockville, MD. If you would prefer to address the PRB via phone, I will also work with you tocoordinate a date/time during the upcoming weeks.

Thank you. Sincerely, John G. LambSenior Licensing Project ManagerNRC/NRR/DORL/LPL1-2(301)-415-3100 

I I .' , , ,

NUCLEAR POWER PLANT SYSTEM SOURCE

.. . '.., ,,..,.l,ll

OYSTER CREEK I

Oysrer Creek

I 1 O F CONTENTS

p&,J

s utxm OON PLANT'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I

ICA7'10N OF SIMILAR KUCLEAK POWER PLASVI'S . . . I

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I \ ~ e r Creek . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

A I ' I I ' A l k l i ~ i r r ~ o n of Syn~bols Used in r t ~ r System and I . ayou~s 1)rawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

, , , . . Oyster Creek

. . . . . : . , ..

LIST OF FIGUR,ES (continued) . . . . , , , I , , . .

3.10-1 Oyster Creek Reactor Shutdown Cooling System ......................... 3.10-2 Oyster Creek Reactor Shutdown Cooling System Showing Component ....................................................................... Locations 4- 1 General View of Oyster Creek ~uc lear Power Plant and

........................................................................ Vicinity 39 ............................................ ster Creek General Site Plan.. 40 , :,

. . . . . . . ,.. ...................... Building Elevation Drawing.. 4 1 .................... Oyster Creek Reactor Building - Elevation (-) 19'-6" 42

.... 4-5 Oyster Creek Reactor Building - Elevation (-) 6-5" and (-) 1'-11". 43 ........................ 4-6 Oyster Creek Reactor Building - Elevation 23'-6" 44

4-7 Oy5,er Creek Reactor Building - Elevations 33'-5" to 38'-0" ............ 4-8 Oyster Creek Reactor Building -,Elevation . . 51'-3" ........................ . . 45 4-9 Oyster Creek Reactor ~ u i l d i n ~ - Elevation 75'-3" ........................ 46 4-10 Oyster Creek Reactor Building - Elevation 95'-3" ........................ 47 4-1 1 Oystcr Creek Reactor Building - Elevation 119-3" ....................... 48 4-12 Oyster Creek Turbine Building - Elevations 0 ' -0 and 3'-6 ............ 49 4-13 Oyster Creek Turbine and Office Buildings - Elevations 23'4

' , .......................................................................... to 27 '0 50 4- 14 Oyster Creek Turbine and Office Buildings - Elevation 35'-0.. ........ 5 1 4-15 Oyster Creek Turbine and Office Buildings - Elevation 46-6". ......... 52 4-16 Oyster Creek Turbine and Office Buildings - Roof Elevations ..........

- ........................... 4-17 Oyster Creek Intake Structure Elevation 6-0 53 4-18 Oyster Creek Freshwatcr Pumphouse - Elevation 1 2 - 0 ................ 4-19 Oystcr Creek Emergency Diesel Generator Building, Elevation 23'0 . 54

Oyster Creek

LIST O F FIGURES (continued)

E l u c e;ire

A-1 Key to Symbols in Fluid System Drawings ................................

A-2 Key to Symbols in Elecmcal System Drawings ...........................

A-3 Key to Symbols in Facility' Layout Drawings ..................... ....

Oyster Creek

LIST OF TABLES

............. 3- 1 Summary of Oyster Creek Systems Covered in this Repon

3.1 - 1 Oyster Creek Reactor Coolant System Data Summary for Selected Components ........................... ... .......................................

3.2-1 Oyster Creek Isolation Condenser System Data Summary for Selected Components ......... .... ..................................................

3.3-1 Oyster Creek Emergency Core Cooling System Data Summary for Selected ......................................................................

, , , , , .,., . . , , . .,.. ?,,a:.,L,,.i. , ........................

ontainment~~ray System Data Summary for Selected ' , , .. ' ...................................................................

,: iii oyster Creek Elecujc Power , System I t 1 . . Data Summary forselected

, . C o m p o n ~ n t s ....................................................................... , ,,

, , , , , . , ' ,

........ Pamid ~ i ' s t i n ~ of Electrical ~,q&es and Loads at Oyster Creek..

Oyster Creek Conuol Rod Drive Hydraulic System Data Summary for Selected Components.. ................................................... Oyster creek Fire Protection System Data Summary for Selected Components .... ; ............ .., ...............................................

, , ,

Oyster Creek Emergency Service Water System Data Summary / I for Selected Components. ....................................................

Definition of Oyster Creek Building and Location Codes ................ Panid Listing of Components by Location at Oyster Creek .............

Component Type Codes.. ....................................................

Oyster Creek 1 , . . . I

. . . .. ,. . . .

The information in this repon has been developed over an extended period of time based on a site visit, the Final Safety Analysis Report, system and layout drawings, and other published information. To the best of our

ge, it accurately reflects the plant configurati the ion was obtained, however, the information in has independently verifiedB'*:Ule licensee or the NR

This sourcebook will be periodically updated with new andlor replacement pages as appropriate to incorporate additional information on this reactor plant. Technical er