ML19210D554

Forwards Response to NRC 790717 Modified Request for Addl Info Needed for Generic Rept on Bwrs.Info Re Containment Isolation Will Be Sent by 791201

ML19210D554
Person / Time
Site:	Oyster Creek
Issue date:	11/20/1979
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Kane W NRC - TMI-2 BULLETINS & ORDERS TASK FORCE
References
TASK-06-04, TASK-6-4, TASK-RR NUDOCS 7911270296
Download: ML19210D554 (22)

Text

. $

i H- Jersey Central Power & Light Company h

ad .q

" - - mw f(f? g) Madison Avenue at Punch Bow! Road Morristown, New Jersey 07960 (201)455-8200 November 20, 1979 Mr. William Kane Bulletins & Orders Task Force U.S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Bulletins and Orders Task Force Long Term Systems Information

Reference:

(a) Division of Operating Reactors Letter dated 7/17/79, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors.

(b) USNRC Letter to General Electric

" Summary of July 12, 1979 Meeting to Discuss BWR Loss of Feedwater and Small Break IOCA" dated 7/26/79.

(c) Letter fran T. D. Keenan, Chairman, General Electric Operating Plant Owners' Group, to Mr. D. F. Ross, Jr. of the NRC dated 8/17/79, General Electric Report "NEDO-24708, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors".

Enclosure:

(1) Bulletins & Orders Task Force - Long Term Systems Information By letter dated July 17, 1979 (reference a), the NRC had requested that Jersey Central Power & Light Company provide additional information which was deemed necessary to enable the NRC staff to issue a generic report on BWRs.

' Ibis information request was later modified by reference (b) . Part of the information requested was sent to the NRC on August 17, 1979 (Reference c).

Enclosure (1) to this letter provides the remainder of the requested information in accordance with the schedule provided in reference (b) except for the following: the contaimnent isolation information requested by Item I of Attachment 2 to Enclosure 1 of reference (a) is still being compiled, and has not been included in this package; it will be forwarded to you as soon as 1396 294 Jersey Centra' Power & Light Company is a Member of the General 7911270 hc Utattes System

Possible, but no later than December 1, 1979.

If you need any clarification of our responses during your review of the information provided, please contact Mr. Jim Knubel, Supervisor, Nuclear Safety

& Licensing, (201-455-8753) of my staff.

Very truly yours,

$. h Ivan R. Finfr k, J Vice Preside la 1396 295

9 t

4 P

ENCLOSURE 1.

1396 296 s

PLANT oyster creek UNIT (S) 1

. BYPASS CAPACITY Plant Steam Bypass Capacity, % Rated 40 1396 297

PLANT oyster creek SYSTEMS AND COMPONENTS SHARED BETWEEN UNITS PAGE CONTINUED PAGE Single-unit plant check here Lxj and do not complete Shared Between System or Component Units Numbers 1396 298 O

5 -

PLANT Oysta.' Creek UNIT ($) 1 PLANT.$PECIF,ICSYSTEMINFORMATION *

. Page 1. continued on page 2.

General Water Sources Instrumentation and Cortrol Mquency of

• Safety $stsmic Safety Seismic Safety Seismic System and system Classifiestf on Catencrz C1sssification Citerr1 Classif. Category Component Tests 1 RCIC N#* NA NA NA NA NA NA 2 Isolation Condenser Note I t _ Note I g Note i Note 2 7yggy pEFUELING OUTACT

3. HPC; NA NA NA NA NA NA NA

_4 HPCI - MA NA NA NA NA NA NA

5. LPC5 Not, I I I Note i Note i Note 2 EVERY 3 DONTHS

6. LPCI NA NA NA NA NA NA NA

7. ADS Note 1 I NA NA =

Note I Note 2 CVERY REFUELING OUTAGE

8. 3Ry . I , NA a Uhli. L bl MT 41ik L L NA Noto I REFUELING OUTACES

9. RHR (includ{ng - - - - -

shutoown cooling, steam condensing.

suppression pool cooling, une. t a

a It I

m.

a a

i i 1

s,3te i i,a .,eo Note 2 e.or usno Otai N Mus b )

contaltunent spray snodes) a II a

r -

unk m n nm , imm

{' ')

. 1 I a

~

10. 55W (setvice water system) = I a ir a Note 2 unknown tytRT 3 FtWTHS NA g

11. RBCC'W a I CIDSED LOOP NA a "

NA y , 12. CRD$ a I Q 2}

Note 1 It a ' Note 2 ONCE PER DONTH

13. CST - a  !! MA NA a linknown NA N 14. Main Feedwatcr

• a i

• Note 1 II " II "

< 15. Rectreviation a I CIDSED IDOP NA " NA

[

• ~ (Jnknown

• Puii'p/ Motor Cooling

'rM I

Note 2: Original Burns 4 Roe design speelfications calls for Category I design criteria. actual conditions in field Indicate a probable non-seismic I installation.

Note It Safety classification did not exist when Oyster Creek was built, see next seven pages for design t code specifications, information provided may be incomplete. U

,

• Some components of these systems are tested more frequently than indicated.

• • An entry of 'HA" is provided when the question does not apply to a particular system or when the system does not exist at the plant.

Pt. ANT-SPECIPlc SYSTEM INFORMATION Sheet 1 OYSTER CREEK Page 2. continued on prae 3.

COHPONENT/ SUBSYSTEM qiniiTY CROUP SEISMIC RUL%RK5 e

I PLANT R.C. I.26 PLANT R.G. 3.29 NOTE DESIGN SRP 3.2.2 DESICN SRP 3.2.1 1

i STRUCT11RES

1. REACitHL BLDC CLASS I SEISMIC ASME SEC VIII CATEGORT I R.C. 1.29 C.I.0/(V.3.5)

2. DRYNELL, TORUS, VENTS 1962 Edition 4 If CLASS I (SEE) R.C.1.29 C.I.o/(V.3.5) 1270N.5

3. C(MTROL ROOM PANELS 1272N.5 CIASS I I 1271N . R.C.1.29 C.I.n/(V.3.5) 4 SPENT PUEL POOL SEC. VIII CIASS I I R.C. 1.29 C.I.1/(V.3.5)

5. ASA. IEEE. ASTM. N/A VENT STACK AwS, N8PU 4 pep CIASS I NSI (OBE)

STATE 4 LOCAL

6. TUtBINE 8tDG. CLASS II

7. RADWASTE BLDG. CLASS !! NONSEISMIC SEE SRP II.3 CATECORY I

~

(OBE) 2.3 9. INTAKE 4 DISCHARCE CLASS II SEISMIC I IF THIS IS l'LTIMATE HEAT SINK IT SHOULD BE SI

~

l0. SCREEN HOUSE CIASS II R.C. 1.29(13).

& EQUIPMENT kS$$

& 3.2,3 1. REACTI)R VESSEL ASME BPV ASME SPV A CIASS I SECT. !

SEISMIC I Pg. IV.1-1/R.C. 1.29 1.2 2. CRDS Housing SECT. lll C

.A ASME BPY A CLASS I SEISMIC IR.C.

.I.a/(V.3.5) 1.29 (C.2/(V.3.5) m SECT. I - - - - - - - - - -

" 1,2 Supports CE 1.2.3 3. CLASS I SEISMIC I C RV SUPPORTS g5 R.C. l.29 C.2/(V.3.5)

I.2.3 4 PUEL ELEMENTS gE CIASS I SEISMIC I 1.2.3 5. CORE SHROUD N/A CLASS I SEISMIC I a.C. 1.29 C.2/(V.3.5) 1.2.3 6. CORE SUPPORTS SE CIASS I SEISHIC I

}R.C.l.29C.I.b/(V.3.5) 1,2,3 7. STEAM SEPARA1M CE CIASS I SEISHIC I (R.C. 1.29 C.I.b/(V.3.5)

ASME 8 II CLASS I R W. I.29 C.I.b/(V.1.5)

SEISMIC I 4.C. l .29 C.I .b/(V.3.5) l emee - **

i i

I 4

k 4

Sheet 2 OYSTER CREEE PLANT.SPECIP!C SYSTEM INFORMATION Page 3, continued on page 4

- l COMPONENT / SUBSYSTEM QUALITY CROUP SEISHIC RINARES PLANT R.G. 1.26 PLANT R.C. 1.29 NOTE DESIGN SRP 3.2.2 DESIGN SRP 3.2.1 1

1,2, A. STEAM DRYER ASME 8 II N/A CLASS I SEISMIC I R.C. 1.29 C.I b/V.3.3 3

9. P! PING PROM RilACTOR ASMB SI 1961 to first ASME 8PV $1II Cl CIASS I VESSEL TO Ist IsotATION SEISMIC I R.C. 1.29 C.I.a ISO VALVE OLTTSIDE DRYlELL A VALVE EXTERNAL TO DkfwtLL then ANSI 831.1 1965

' 4 NUC Code Cases to 2nd ISO valve.

RECIRCULATION SYSTEM

'1 1. PUMPS ASA B31.1 ASMB SPV SIII Cl A CLASS I ji 2. VALVES ASME BPY I a A SEISHIC 1 VI 1.I/R.C. l.29 C.I.a CIASS I SEISMIC I IV.I /R.C. 1.29 C.I.s

]1 3. PIPING ASME SPV 1 " A CLASS I SEISMIC I IV.I 4/ R.C. 1.29 C. I . A 9 ASA B31.1 EMERCENCY SYSTEMS

1. ISOLATION CONDENSER 2.3 SHELL ASME BPY IV.1 2 CIASS I SEISHIC I R.C. I.29 C.2

! SECT. VIII C

.I,2 3 TUtg ASNE SPV ASME SPV SIII C2 CLASS I SEISMIC I R.C. 1.29 C.I.b e

S.!!! Cl.A B SIII CIASS 2 gl.2 S PIPING ASME Sec. I ASME SPV

~ 1965 $1II CLASS 2 B CIASS I SEISMIC I R.C. 1.29 C.I.b 3' 2. LIQUID P0! SON SYS. ASME 31 1963 W ASME SPV

• OntP if CRDS falls to ANSI 8 31.1 1963 8III CIASS 2 8 CLASS I SEISMIC I R.C. l.29 C.I.b g operste 4 NCC PUMP ASMB SIII CIC B

@ TANK API B l 3 CORE SPRAY SYS. ASME BPV SIII C2 CLASS I SEISMIC I

,1,3 PIPING ASA B31.1 A/B " R.C. I.29 C.I.c/VI.6.2

,4

'" CIASS I

. 1 REACTOR BLDC O 2 CIDSED IEOP COOLING ANSI.331.1 1963 4 ASME BPY g CIASS I i SP.tSMIC I V.3.5/R.C. 1.29 C.I.g

_ Nuclear Code Cases Sill CL 3 1 3 5. AITIUMATIC DEPRESSUR.

l 4 l Part of ECCS; should be i Selsmic 1. Class 2 l 4 I

_ ._ _ f

PLANT-SPECIPIC SYSTEM INFORMATION Page 4. continued on page 5 Sheet 3

/.

• OYSTER CREEK CMPWENT/SUBSYSTD4 QUALITY GROUP SEISMIC REMARKS PLANT R.C. 1.26 NOTE PLANT R.C. 1.29 I DESIQi SRP 3.2.2 DESIGN SRP 3.2.1 .

REACTOR DERGENCY SYS.

ASME BPY 3

SERVICE NATER SYSTEM SI 4 II ASMB SPV CIASS I SEISMIC I To ist 150- S!!I CIASS 3 C R.G. 3.29 C.I.3/I-3-2 g IATION VALVE REST ASA 1331.I-1955* .

CHANGES TO 1966 3 CONTAINMENT SPRAY ASA 531.1 ASME SPV CLASS I SYSTEM 8EISHIC 1 R.C. 1.29 C.I.c/VI-7-1 ASME S VIII S!!! CIASS 2 8 10 1st VALV8 OtrTSIDE C0;tTAlfelENT I STAN08Y CAS TREAT. SYS CE 8** CIASS I SLISMIC l' No description of system

• 1f system removes H3 .

• • no reg't if post sceld fission ASTM. A373 product removal is not basis.

~ SPENT FUEL AND NEN A245, 6VA 81C 0.C. uses H2 Inerting system I FUEL STORACE PACILITIES ASTM, A240 Cl g CIASS I SEISMIC Puel Handling Not Covered by R. G.

304L Q I.26lR.G. 129 C.I.L requires selsmic

@ DIERGENCY ELEC. SYS. ASA. ASTM. ASTN . CLASS I SEISMIC 1 R.C.1.29 (C.I.q) CIASS IE ONSIT11 1.3 RATTERIES IEEE NEMA.NFB J POWER SUPPLIES IEE8 344. SRP 8.3.2 (M 2.3 DIESEL CENERATOR REMA.14 CAL ASME.ASA.ASTN,  !

Q IEEE. NEMA.NPSJ i R.G. 1.29 C.I.q Class II thsite Power DEMI.14 CAL l

' 2. 3 ASA.!EEEe EMERCENCY SUSSES. J ETC. ASTM, NEMA R.G. 1.29 C.I.8 Class 18 Onsite Power SUPPL!ss IEEE 344. SRP 8.3.1

I.

I FIANT-SPECIFIC SYSTEM INFORMATION Sheet 4 Page 5, continued on page 6 OYSTER CREEE CNPONENT/ SUBSYSTEM QUALITY CROUP SPISMIC PEMARKS 4

t FLANT R.C. 1.26 FIANT R.C. 1.29 NOTT DESIGN SRP 3.2.2 DESIGN SRP 3.2.8 I

l STEAN SYSTEM

'! 1. Flying to Ist.l.V. ASMB BPV ASME BPY CIASS 1 SEISMIC I R.C. 1.29 C.I.e/IV 1 2 8ECT. I A III CIASS I A

2. Flping to Turb SV ASA B31.1 ASME BPY SEISN!C I R.C. 1.29 C.I.o SIII CIASS 2 9

,l1 3. MSIV VALV8 ASMB SPV ASMB BPY CIASS I SBISMIC 1 R.C. 1.29 C.I.a/IV-1-2 i SECT. L SIII CLASS 1 A ASA 831.1 G8 SPECS

4. TURBINE SMP YALV8 ANSI B31.1.1.0, D NON SEISMIC I SRP 3.2.1 m 1 S. SAFETY VALYB ASME BPV ASM8 BPV CLASS I S8ISMIC 1 R.C. 1.29 C.I.e/IV-1-2 SECT. I SIII CtASS I A U RNCC-1271N

.I 6. RELIEF VALY8 ASNB BPV ASME BPV CIASS I fg

• SEISMIC I R.C. I.29 C.I.a/IV-t-2 SECT I Sill CIASS 1 A I

y __ ,! SHUTDOWN COOLING ---m--- ------- CLASS II --------- I-2-2 SYSTD45 ASMB BPV ASME SPV CIASS II SEISMIC I R.C. I.29-C.I.d O 2 PUMPS SECT.III SIII CIASS 2 C u

SIAJTDOWN COOLING ASMB BPV ASMS BPV CIASS II SEISMIC I R.C. I.29 Col.d 2 TUS8 SIIT SIII CIASS 2 C CLASS C

- s e

PtANT-SPECIP!C SYSTEM INFORMATION Sheet 5 Page 6, continued on page 7 j

, OTSTER CREEK t CD(PONENT/SUBSTSTEM QUALITT CROUP SEISMIC RDtARKS PIANT R.C. 1.26 PLAAT R.G. 1.29 NOTE DESIGN SRP S.2.2 DES!Qt SRP S.2.1 1

2 x SHELL ASMB BPV ASME BPY C CLASS II SBISHIC I R.C. I.29 C.I.g SECT. VIII SIII CIASS 2 SIII CIASS 2 CONDENSATE STORACE CIASS II SEISMIC II CONDENSATE STORACE SYSTD4 IS TANKS 4 PtNPS Analysed for REIATED TO S!,FETY; ECCS SMALL O.113 selss!< EREAKS USE CRD WH!Of HAS NORMAL event - 0 K. SUCTION FRO 4 CST ISOLATION CONDENSER CAN ALSO TAKE SUCTION FROM CST.

I LIQUID WASTE ST8 TEM API 650 ASME BPY D CtASS II WASTE MANACD4ENT SYSTDtS IS STANDARD SIII CIASS S UNDER DEVELDPMENT SRP 5.4.8 REQUIRES TitAT PIPING FROM THE RECIRCUtATION IDOP TO Tile (<

^ g/

OUTERMOST IsotATION VALVE BE

~

CLASS A SEISMIC I g

{

X REACTOR CLEANUP STS. ASME BPY ASME BPY CIASS II NQt-SEISMIC I-2-1 O 4 SIII SIII CIASS S C W CIASS C

'K AIR C04 PRESSOR 4 NOT CUVERED CIASS II NON-SEISMIC AIR OPERATED VALVES IMPORTANT TD h U REC 3ItBR3 BT R.C. 1.26 SAFETY ARE REQUIRED TO HAVE Q SEISMIC ACCUHJtATORS 2,3 STATION AUIILIART SUSSES N/A CIASS II g

I McISTURE SEPARATORS AND e REHEATERS SEC. N/A CLASS II g NOTE I = fl denotes equipment required to, hold together to prevent e (DCA due to the setaale event

1. 2 denotes equipment required to shutdown 4 hold in safe shutdown condition following seismic event

1. 3 denotes equ!Pnent required if plant experiences e IDCA '

Ic U

1 denotes equipment not needed for safety during seismic event

• - ~ ~ - ' ~

PLANT. SPECIFIC SYSTEM INFORMATION rage T continued on page 8

0) STER CREEK Sheet 6 C0'IPONENT/ SUBSYSTEM QtL1LITY GROUP SEISMIC RfluRES W PLANT R.C. 3.26 PLANT R.C. 1.2g DESIGN SRP 3.2.2 DESICN SAP 3.2.l 1

INSTRLNENT AND Q)NTROL 2, 3 2, 3 REACTOR LEVEL INSTR. 338.1 SI 4 Y! CIASS I SEISMIC I R.C. 1.29 C.I.k/V.3.6 FEED WATER CONTROL VALVES ASA, IEEE, ASME, ASTM, ANS, NBPU, " " "

4 STATE 4 LOCAL 2' I

, LIQUID POISCN SYS. INSTR. " " "

lANUAL REACTOR CNTROL N/A a a ASA 4 IEEE n 2e 3 CONTROL R00 INST #

• n 2, 3 CONT. ROD POSITION a n n i!NDICATING SYST. a -

a n 2, 3 i REAcr0R PROTECT!m SYSTEM a n a 2e 3 [NElrTRON M3dITUR SYSTEM ASHE III Cl.A a n , n FUEL RUPTURE DETECTIM SYST.' Pressure Parts a n n

>REA MMITVRS CMS $I ......... ..... ..........

1 lTURBINECENERATOR ASME, ASA, ASTM, Class II No Seismic Req't/V 3 6 IEEE, NEMA, NBPU f \

STDS OP TUBULAR ~

E101 ANGER MANU.

FACTUREPS A550-I h h CIATIm CIASS R. N/A No Selsmic Req't/V.3 6 l

• NJ 4 LOCAL CODES h h

~

~

CONDENSER HEAT E101 ANGER Dyr.amic analysts

_ g INSTITUTE N/A Class II for 0.113 seismic e event W FEEDWATER SYSTEM ASMB $1 1963 g ,ltEATERS RV TO !st VALVE * **The FWS is used as re.

IPtNPS OlfrSIDE CONTAIN. CIASS Ig dundant ECCS to the ADS for Small Brks

@ ' MDIT , and single fallure requiressats. This ASME SECT, VIII i would require that the FWS sel.ste I, Amend. 38 pres nts results of seismic 3 tw l TFMA STDS. , C/

analysis to sh that the feedwater in.

jection systes meets Class I standards. } j S

b i

I i

PIANT-SPECIFIC SYSTEH INFORMATION Page 8, FINAL. Sheet 7 QUALITY STANDARDS t

1

, NRC QUALITY CROUP A NRC QUALITY CROUP 5 NRC QUALITY CROUP C NRC QUALITY CROUP D ASME Soller and Pressure ASME, Sec. VIII Vessel Code,Section III (pressure vessels)

ANSI 838.1.0

+

Class 1 Class 2 Class 3 (plying and valves)

Nfg. Stds. (pumps) or or ASME Bo!!er and Pressure Vessel Code,Section III Class A Class C .. ..

(pressure vessels) (pressure vessels)

W W NRC Quality Group A equivalent to Licensee Safety Clase I w

g ) Also seisale Category I NRC Quality Croup B equivalent to Licensee Safety Catss 2 NRC Quality Croup C equivalent to Licensee Safety Class S } Not etways seismic Category I NRC Quality Croup D equivalent to Licensee Safety Class 4 or NNS } Non-seismic Category I o ,

i

[hs c3 Y

2EED e

PLANT: 0YSTER CREEK NUCLEAR STATION UNIT 1 Page 1, continued DESIGN REQUIREMENTS FOR CONTAINMENT ISOLATION BARRIERS on page 2 Question: Discuss the extent to which the quality standards and seismic design classification of the containment isolation provisions follow the recommendations of Regulatory Guides 1.26, " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Water-Containing Components of Nuclear Power Plants", and 1.29, " Seismic Design Classification".

Response: The table shown below lists quality standards and seismic design classification used fcr the containment isolation provisions for Oyster Creek Nuclear Station. Also listed in the table are the standards and classifications recommended by Regulatory Guide 1.26 and 1.29.

QUALITY GROUP SEISMIC DESIGN SYSTEM PLANT DESIGN R.G. 1.26 PLANT DESIGN R.G. 1.29 Piping From Re- ASME BPV ASME BPV Class I Category I actor Vessel to Section I Section III 1st Isolation (1965) C 1, A Valve external to Drywell Recirculation ASME BPV ASME BPV Class I Category I line Section 1 Section III ASA B31.1 C1, A Isolation Con- ASME BPV ASME BPV Class I Category I denser Section 1 Section III (1965) C2, B Core Spray AS:1E BPV ASME BPV Class I Category I Section I Section III ASA B31.1 C2, A/B Containment ASME BPV ASME BPV Class I Category I Spray Section VIII Section III C2, 8 Steam Line ASME BPV ASME BPV Class I Category I Section I 3ection III ASA B31.1 C1, A GE Spec.

Reactor Clean ASME BPV ASME BPV Class II Non-Seismic Up Section III Section III Class C C3, C 1396 307

DESIGN REQUIREMENTS FOR CONTAINbENT ISOLATION BARRIERS Page 2, FINAL QUALITY GROUP SEISMIC DESIGN SYSTEM PLANT DESIGN R.G. 1.26 PLANT DESIGN R.G. 1.29 Feedwater System ASME BPV D Class II Category II Section I (1965) .

Amendment 38 If the FWS to the Oyster is used as Creek FSAR redundant provides result s ECCS to the of seismic anal - ADS for ysis to show small breaks that the and single feedwater in- failure jection system requirements, meets Class I it would standards- require that the FWS be Category I, 1396 308

PLANT Oyster creek UNIT (S)1 PROVISIONS FOR TESTING Oue: tion: Discuss the design provisions for testing the operability of the isolation valves.

Response: The operability of primary containment isolation and reactor coolant ,

isolation valves is tested during every refueling outage. Testing of primary containment isolation valve operability is performed ,

in two ways: (1) by introducing a high drywell pressure signal to the drywell high pressure sensor and (2) by introducing a Lo Lo water level signal to the Yarway 1cvel sensors. Testing of reactor coolant isolation valve operability is performed by introducing a Lo Lo Water IcVel signal to the Yarway 1cvel sensors.

Testing is performed by the following steps:

(1) Isolate the pressure sensor or the Yarway level sensor to be tested.

(2) Introduce an appropriate signal (either a high pressure signal or a level d/p signal) to the pressure sensor or level sensor until the appropriate relays trip.

(3) Verify that all relays function as specified.

(4) Restore the pressure sensor or level sensor back to service.

(5) Repeat steps (1) through is) until all sensoes are tested, step (4) is omitted for the last sensor testud in order to obtain a 2 out of 4 coincidence trip wher. step (6) is performed.

(6) Introduce an appropriate trip signal to one of the other sensors previously tested to obtain a 2 out of 4 coincidence trip.

(7) Verify that all isolation valves are closert in accordanca with procedure requirements.

(8) Restore all sensors and other system components back to service.

(9) Individually close and record closing timt s of the reactor coolant isolation valves.

These tests are considered acceptable if all requirements specified in the appropriate procedures are met.

1396 309

PLANT Oyster Creek UNIT (S) 1 CODES, STANDARDS, AND GUIDES Question: Identify the codes, standards, and guides applied in the design of the containment isolation system and system components.

Response: 1) Piping A) From vessel to first - ASME Sec. I - 1965 Isolation Valve B) After first ISO valve - ANSI B31.1 - 1965 to outside ISO valve

2) Valves and Operators - Revisions to 1965 American Standards Association American Society for Testing bhterials ASME - Section I and VIII Pipe Fabrication Institute American Institute of Electrical Engineers National Electrical Manufacturers Association

3) Instrumentation Except for material used in fabrication of instrumentation, manufacturer's standards apply.

1396 3i0

PLANT Oyster Creek UNIT (S) 1 NORMAL OPERATING MODES AND ISOLATION MODES Question: Discuss the normal operating modes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out of the containment.

Response

Main Steam Isolation Valves Each main steam line is provided with two isolation valves, one inside containment and one outside, which are open during normal operation. A test circuit is provided to permit slow speed exercising of each valve.

Automatic closure of the valves is initiated by the following signals:

e

a. High flow in the main steam line.

b. Rx Lo-Lo water level.

c. Steam line high radiation.

d. High temperature in steam tunnel.

e. Low pressure in main steam line (82Spsig).

Main Steam line drains are closed in normal operation. These valves are (automatically) closed by the same signals associated with Main Steam isolation valves.

Energy Condenser Valves The system is operable and ready for service at all times during power operation. Conditions requiring use of the system are considered abnormal. The two series valves in the steam inlet lines are open and one of two series valves in the condensate return line is open during normal operation. The vent isolation valves are normally open.

Automatic opening of the condensate return line is initiated by the following signals:

a. Rx Lo-Lo water level,

b. Rx high pressure.

Isolation of the steam inlet line and condensate return line is initiated by excess flow in either line.

1396 311

Response: (Continued)

Feedwater System Feedwater lines connected to the reactor vessel do not utilize automatic isolation valves, since operation of this system is essential following a loss of coolant accident. Check valves located in these lines inside the drywell and outside the drywell provide isolation when necessary.

Liquid Poison System The liquid poison system is a stand-by system for use in the event that the control rod system is inoperable. The system is activated only by remote manual action from the control room; that is the squib valves are fired and pump started.

Isolation is providedby two check valves.

Shutdown Cooling System The shutdown cooling system removes fission product decay heat during shutdown with reactor pressure less than 150 psi. The isolation valves inside the drywell are intericcked to prevent system operation above 3500F.

Isolation valves in the suction and discharge headers are auto-matica11y closed by the following initialing signals:

a. Rx Lo-Lo water level.

b. High inlet temperature (3500F).

Clean Up System The clean up demineralizer system is a filtratior and ion exchange system for maintaining the purity of the water in the reactor vessel and is normally in service with isolation valves open. The influent piping has isolation valves inside and outside the drywell. The return pipe has an isolation valve outside drywell and a check valve inside the drywell.

Automatic isolation valve closure is initiated by any of the following conditionr:

a. Rx Lo-Lo water level.

b. Low flow through the clean up filter.

c. High temperature Rx water from non-regenative heat exchanger.

1396 312

Response: (Continued)

Clean Up System (Continued)

d. High pressure from the pressure reducing station.

e. High temperature cooling water from the auxiliary clean-up pump.

f. Liquid poison system flow to vessel.

An exception to the above is the return valve V-16-61 which closes on lo-lo reactor water level only.

Core Spray System The core spray system is a low pressure stand-by core cooling system which provides an alternate supply of reactor cooling water in the event of a pipe break accident in the primary system. The isolation parallel valves are normally closed and do not open until after the system initiates from Rx lo-lo water level or high drywell pressure and reactor pressure is less than 285 psi.

Check valves are provided in each header inside drywell which provide isolation. The manual isolation valves inside the drywell are locked open and all other system valves in the flow path are open. The valves which recirculate water to the torus are closed except during testing.

Control-Rod Drive System The control rod hydraulic system has two isolation valves.

'!he first if a ball check which is part of the control drive mechanism u.d a normally closed hydraulic system control valve.

The exhaust header takes water discharged by the drives during operation and returns this water to the reactor check valves and effectively isolates this line.

Instrument Air / Nitrogen Control air and nitrogen valves are open for control instrument use-during power operation. No automatic isolation is presently provided.

Drywell 6 Torus 07 Sample System The isolation valves in this system are open during power oper-stion.

1396 313

Response: (Continued)

Drywell 6 Torus 02 Sample System (Continued)

Isolation is provided automatically from the following signals:

a. Rx Lo-Lo water level.

b. Drywell high pressure.

Rx Building Closed Cooling Water System The closed cooling water lines do not connect to the reactor primary system and are not open to the containment atmosphere.

They are provided with isolation valves in the supply and re-turn headers. No automatic isolation provided.

Reactor Primary-System Instrument Lines Reactor primary-system instrumentation piping which leaves the primary containment are dead-ended at devices' location in the reactor building. These lines are provided with manual isolation valves outside drywell and excess flow checks in each line which close automatically with leakage flow.

Containment Spray System The containment spray system is not provided with automatic isolation valves, but does have motor operated isolation valve that can be operated from the Control Room.

Reactor Head Cooling System The head cooling system is used in conjunction with reactor vessel flooding and shutdown cooling system for vessel cool down. Manual and motor operated isolation valves are closed during power oper-ation. No automatic isolation is required.

Drywell Equipment Drain Tank (DEDT) S Sump Drain Floor and ec,uipment drains are segregated in the drywell with the floor drains collected in a sump and the equipment drains collected in a drain tank. The contents of each are transferred to Radwaste Building. The DEDT tank is provided with overflow and vent lines to the sump.

Each discharger has dual isolation valves that are closed auto-matica11y by the following signals:

a. Rx Lo-Lo water level.

b. High drywell pressure. l 3h[j j.l4

Response: (Continued)

Drywell Purge S Exhaust System During reactor operation, the nitrogen atmorphere is isolated.

within the drywell and recirculated by the drywell cooler blowers. The make up is automatic on pressure control to the drywell only. In each supply line, drywell and torus series isolation valves are closed during reactor operation to isolate the containment vessel from the make up line, except when make up is required. All isolation valves are automatically closed by the following signals:

a. Rx Lo-Lo water level,

b. High drywell pressure.

A keylock bypass switch is installed which bypasses any dry-well isolation signal allowing the operator to open any of the drywell purge as vent valves. The exhaust is vented normally to the stack with isolation valve normally closed and interlocked to close automatically by the following signals:

a. Rx Lo-Lo water level.

b. High drywell pressure.

All operable isolation valves in the containment inerting system are automatically closed. A keylock bypass switch is provided in the exhaust line to the gas treatment system to reduce pressure in the torus and drywell in the post accident phase.

Vacuum Relief System Automatic vacuum-relief devices are used to prevent the primary containment from exceeding the external design pressure.

The drywell vacuum-relief device draws atmosphere from the pressure absorption chamber and the pressuro absorption chamber vacuum-relief device draws air from the reactor building. Two vacuum breakers in series are used in each of the two lines to atmosphere. One valve is activated by a differential pressure signal while the second valve is a check valve.

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