ML20101E314

Forwards List of Environ Qualification Equipment Replaced During Previous Outages & Justification for Continued Operation,Per NRC 840509 Request.No Significant Degradation of Safety Function Will Occur During Design Basis Event

ML20101E314
Person / Time
Site:	Oyster Creek
Issue date:	12/21/1984
From:	Wilson R GENERAL PUBLIC UTILITIES CORP.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
NUDOCS 8412260208
Download: ML20101E314 (58)

Text

M g

GPU Nuclear Corporation

'~

RUOEME 100 interpace Parkway Parsippany, New Jersey 07054-1149 (201)263-6500 TELEX 136-482 Writer's Direct Dial Number:

December 21, 1984 Mr. John A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C.

20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Environmental Qualification of Equipment Important to Safety Your letter dated May 9, 1984 requested GPUN to provide additional information which the NRC staff requires to complete their Safety Evaluation Report on environmental qualification for Oyster Creek Nuclear Generating Station (0CNGS).

It also requested certification that certain efforts had been conducted.

It is the intent of this letter to provide the information and certifications.

Our letters dated June 15, 1984 and September 14, 1984 requested additional time to respond to the information required by the May 9, 1984 letter. To this end GPUN established a new environmental qualification (EQ) organization and program to review the entire scope of the EQ efforts for OCNGS.

The program has resulted in a detailed re-examination of the previously submitted (Environmental Qualification Report dated Nov. 1, 1980, GPUN letters dated March 16, 1983 and March 16,1984) environmental qualification equipment master list (EQEML) and justifications for continued operation (JCO's).

In the course of the re-examination, the Oyster Creek EQEML has been revised. Components which are believed to fall within the scope of 10CFR50.49 have been added, and other components have been deleted. The EQEML was developed based on reviews of Oyster Creek documents such as Facility Description and Safety Analysis Report, Technical Specifications, Emergency Operating Procedures, Piping and Instrumentation Diagrams, electrical distribution diagrams and System Design Description. A GPUN Technical Data Report (TDR) which describes equipment locations and environments has also been used.

It should be noted that the TDR is being revised as a result of this review. The revised EQEML and the report documenting the methodology utilized to g'nerate the EQEML are available in GPUN's corporate office. The h

8412260200 041221

,0i PDR ADOCK 05000219 e1 P

PDR

\\

GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation 1

t

=

' s}

~, +

z

~

l u,

m -

m m+

'following discussion and the attachments 'to this letter provide our

- t "V

resp nse to the May 9. 1984 letter.

'The ievised EQEML includes equipment that is relied upon to remain functional during~and following design basis events to ensure the-integrity of the reactor coolant pressure boundary, the capability to

.' achieve and maintain a safe shutdown condition.,and the capability to prevert;or mitigate the consequences of accidents.

t 7

'All design ~ basis events which could potentially result in a harsh o

nenvirotment, including flooding outside containment, were considered in the devslopment'of the revised E.QEML.

.In addition to the above, the following were considered and factored into the EQEML list:

1 1.

Auxiliary devices electrically connected directly into the control or power circuitry of the safety-related equipment whose failure due to postulated environmental conditions could prevent required operation of safety-related equipment.-

2.~

Mechanically connected auxiliary systems with electrical

. components which are necessary for the required operation of safety-related equipment.

3.

Post accident monitoring equipment as required by Regulatory l

' Guide 1.97(Rev.2).

W. components shown.on-the EQEML are being verified by our extensive iteld verification in terms of their. identification (tag number),

location, model and manufacturer to ensure that our documentation reflects the actual installation.

Item 2d of the NRC's May 9, 1984 letter requested GPUN to address non-safety electrical circuits indirectly associated with the electrical equipment by common power supply or physical proximity.

Those non-safety circuits which are related by " common power supply" are being addressed by a review of :Se plant electrical design to verify that all such circuits are protected by properly coordinated protective

' devices which will ensure that failure of a non-safety related circuit will not cause loss of a power supply to qualified electrical equipment, Those non-safety circuits which are related by " physical proximity" of their wiring are addressed by the fact that the plant design standard is that all circuits are protected by fuses or circuit breakers which are

- properly sized to protect the circuit wiring. This insures that damage due to faults will be limited and will not result in either fire or excessive heat'in raceways or enclosures which might disable qualified electrical equipment. As an extension of the electrical design review noted above, a review is being performed to verify that properly sized fuses have been specified and installed.

i

..E....

...m

Q i y --

x; w

~

-3.

Non-safety related" circuits related by "physica1' proximity" of equipment are much,less likely to be sources of damage to qualified equipment.

There is no mechanism by which environmental conditions would cause a 1

-spontaneous explosion in an electrical device. Heat generation due to circuit faults will remain localized and restricted to a short duration.

~because of the' energy limiting effects-of the protective devices referred

'to above.

}

I JC0's'which were previously submitted have been re-examined during the course of our review. As a result, some JCO's have been upgraded, others have been deleted and some have been added. The reason for the deletions iis that some components have been qualified or replaced with qualified l

equipment. Attact. ment I to this letter provides the revised JC0's which supersede the previous JCO's. The JCO's conclude that no significant

' degradation of any safety function or misleading information to the operator will occur as a result of failure of. equipment under the accident environment resulting from a design basis event. Schedules and

. resolution (qualification or replacement) are also provided in'the i

JCO's. The components which have been identified recently, and added to the EQEML, are currently being evaluated for their qualification.

If a component is found not to be qualified, a new JC0 will be generated. Any new JCO's will be submitted to you by January 31. 1985. Althoegh this letter does not constitute'a formal exemption request from the modification schedule as provided in 10CFR50.49, it is anticipated that 1

. exemptions for certain pieces of equipment are required. Following our ongoing review of the qualification status, our exemption request will be

. submitted during the month.of January 1985.

ATTACHMENT-II to this letter identifies that equipment on the EQEML which has already been replaced with qualified equipment in the previous outages through our environmental qualification efforts. As shown_in this attachment, a number of replacement tasks have already been

' accomplished. GPUN will continue to maintain our expanded effort to

-complete the remaining environmental qualification tasks.

y truly yours, Vice President Technical Functions 1r/0493e l

cc: Administrator i

Region I~

U.S. Nuclear Regulatory Commission 631 Park Avenue

(

King of Prussia, Pa.

19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N. J.

08731 l

l i

r ATTACINENT I OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)

JCO-0C-84-1, Rev. O December 19, 1984 Page 1 of 3 a

I 1

l OYSTER CREEK NUCLEAR GENERATING STATION f-L JUSTIFICATIONFORCONTINUEDOPERATION(JCO) 9 i

i i

4 f

JCO-0C-84-1, Rev. O December 19, 1984 Page 2 of 3 COMPONENTS Tag Numbers IA-83A, IA-83B, IA-83C, IA-83D, IA-83E Description Pressure Switches to open EMRVs on high RPV pressure OBJECTIVE The objective of this discussion is to determine:

that the safety function can be accomplished by some other qualified components; that the failure of the identified components as a result of a harsh environment will not degrade other safety functions or mislead the ope rator.

COMPONENT LOCATION All of these switches are located in the Reactor Building on elevation 51'3".

Pressure switches IA-83A, B are located on the east drywell wall; switch IA-83C is located in the southeast quadrant; and switches IA-83D, E are located in the northwest quadrant.

COMPONENT FUNCTION These components open the EMRVs on high RPV pressure to provide protection against over-pressurisation. These switches would be required to work only for a small break LOCA which is not large enough to remove decay heat and when both isolation condensers are not available.

EVALUATION For breaks inside the containment, these components are in a mild environment and will function properly. For breaks outside containment, these switches are located in different areas in the Reactor Building such that the areas will not be in the same harsh environment simultaneously.

In addition, there are 16 safety valves on the steam line which would provide over-pressure protection for the RPV in the event that the EMRVs are unavailable or insufficient to relieve the increasing RPV pressure. Further, the operator can manually operate the EMRVs to depressurize the RPV.

The failure of these switches will not affect the operation of the ADS nor mislead the operator, since the operator is instructed in the E0Ps to manually operate the EMRVs if they fall to initiate automatically.

If a single isolation condenser is available, over-pressurization will not occur and the switches are not required.

JCO-0C-84-1, R2v. O December 19, 1984 Page 3 of 3 CONCLUSION

.The failure of the pressure switches will not prevent over-pressure protection of the RPV and operation of the EMRVs.

The operator will not be misled by this condition.

QUALIFICATION PLAN The switches will be replaced with qualified switches during the November 1985 outage.

O e

i

r. -

JCO-0C-84-2, Rev. O December 19, 1984 Page 1 of 2 l'

'd OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) m N

.a l

l I

l L-

B

~

7 f

y 1JCO-0C-84-2, Rev."O~

December 19..1984

-- s

.lf Page 2:ofl2

COMPONENTS'

,s~

^

. Tag Numbers-V-21-5, V-21-11:

Description-1 Containment Spray Drywell'= Injection Valve Motor' Operators.

t OBJECTIVE 1

5 iThc'.cbjective'of this discussion is~to determine:

~

=that there is a basis for concluding that the~ existing' components will perform their required function; that the~ failure of these components will not degrade-other safety-

~

w functions nor mislead the operator.-

s

.k

. COMPONENT LOCATION These valves'and their associated operators are located in the Reactor Building on the 23'6" elevation.. V-21-5 is. located in the Southeast quadrant-and V-21-11 is located in-the northeast quadrant.

K COMPONENT FUNCTION-These' valves allow containment spray injection into the drywell to remove heat-and reduce containment pressure for breaks inside the drywell.

y : EVALUATION =

The valves and their associated operators are located in the Reactor Building.

,and thus are'not subject to the harsh environmentsinside the drywell when they

~

1are required to function. The environment in the' Reactor Building is not-expected to become harsh due to a break inside the containment, and hence these components are expected'to function. For breaks outside containment, the Containment Spray' System is not required to function to mitigate the These valves are normally open and are not required to change position o

• -event.

xcept during the monthly surveillance testing of the Containment Spray

~

e During this test, only'one of the two Containment Spray Subsystems is System._

tested at any-one time. The operable subsystem provides ample capability to satisfy the function of the system, h

-The operator in the control room has position indicating lights for these

. valves and thus, the failure of these valves would not mislead the operator.

CONCLUSION The failure of one of these valves to operate does not degrade the effective-L ness of the Containment Spray System to perform its function.

- QUALIFICATION PLAN LAn: evaluation will be performed to confirm that these components are not they exposed to a harsh environment'in the Reactor Building for the accident

&re required ~to mitigate, during the time they are required to function. Any unqualified components in the motor operators will be replaced with qualifled components.by November 1985.

~

.g t\\.

n s

i JCO-0C-84-3, Rev. 0 December 19.-1984 Page 1 of 2 s

\\

f.

\\

3 3

4 i

s

\\

il N

c OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 3 g%i b

iz e

4

v

- :q; L

m

.( - M L "

JCO-0C-84-3 -Rev. O

~

December 19, 1984 Page'2'of 2-1 COMPONENTS

Tag Numbers =

V-5-167. V-5-147 Description-RBCCW1 Containment Isolation Valve Motor Operators-

~ '

N

' OBJECTIVE-The objective of this discussion'is to determine:

V4 ' ihatthefhilureof'thesecomponents(V-5-167,147) will'not degrade t

, other safety functions;

that-the' safety function can be accomplished by some other qualified'

-equipment.

9 COMPONENT LOCATION 7.y q

'Both of these valves and their associated motor. operators are located in the southeast. quadrant in the Reactor. Building on the 23'6" elevation.

n

- q>.

' COMPONENT FUNCTION-

.These components function to isolate the drywell upon a containment isolation

~

A, signal resulting from a design basis accident.

,3 -

EVALUATION-Both components are located outs'ide the drywell and are not' expected-to be affected by inside containment breaks.

In the event of a break inside the drywell, both valves will close upon a receipt of a containment isolation signal.

4

- For. breaks outside containment, V-5-167 is in series with a redundant qualified isolation valve (V-5-166)-inside containment which would function for tho'se breaks-(outside' containment) which could create a harsh environment near V-5-167.

The inlet RBCCW valve (V-5-147) is in series with a check valve (V-5-165)J1ocated inside containment'which would function to prevent releases from containment in the event of a break (outside the drywell) which creates a harsh environment near V-5-147.

i s

' CONCLUSION

-The failure of these components will not degrade the isolation or containment since both are in series with components which will function to isolate the drywell;in the event of a DBA.

' QUALIFICATION PLAN An evaluation will be performed to confirm that these components are not exposed'to'a harsh environment in the Reactor Building for the accident they are required to mitigate, during the time they are required to function. Any unqualified components in the motor operators will be replaced with qualified components by November 1985.

1 JCO-0C-84-4, Rev. O December 19, 1984 Page-1 of 3 G

l OYSTER CREEK NUCLEAR GENERATING STATION

' JUSTIFICATIONFORCONTINUEDOPERATION(JCO)

/

r e

1 4

b.'

T JCO-0C-84-4, Rev. O December 19, 1984 Page 2 of 3 np-COMPONENTS

.~

Tag Numbers' V-21-13, V-21-15, V-21-17, V-21-18 Description V-21-13, 17 -Containment Spray Dynamic Test Valve Motor Operators V-21-15, 18 -Containment Spray Torus Spray Valve Motor Operators OBJECTIVE The objective of this discussion is to determine:

that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade other safety functions nor mislead the operator.

COMPONENT LOCATION These valves and their associated motor operators are located in the Reactor Building on the 23'6" elevation.

V-21-13, 15 are located in the southeast quadrant, and V-21-17, 18 are located in the northeast quadrant.

~ COMPONENT FUNCTION These valves allow containment spray injection into the torus, to remove heat and reduce pressure for breaks inside the drywell. The V-21-15, 18 valves are used in the normal mode of containment spray to provide a small spray flow to the torus. The V-21-13, 17 valves are used in the dynamic test mode.to provide cooling to the torus pool.

EVALUATION These valves and their associated operators are located in the Reactor Building and are not subject to the harsh environment inside the drywell when they are required to function. The environment in the Reactor Building is not expected to become harsh due to a break inside the containment, and hence these components are expected to function. For breaks outside the drywell, the Containment Spray System is not required to function to mitigate the event. These valves are in their desired accident position normally except during the monthly surveillance testing of'the Containment Spray System.

During this test, only one of the two Containment Spray Subsystems is tested at any one time. The operable subsystem provides ample capability to satisfy the function of the system.

The operator in the control room has indication of valve position and thus, the failure of these valves would not mislead the operator.

CONCLUSION The failure of these valves to operate does not degrade the effectiveness of the Containment Spray System to perform its function.

JCO-OC-84-4, Rev. O December 19, 1984 Page 3 of 3 QUALIFICATION PLAN An evaluation will be performed to confirm that these compon/ nits are not exposed to a harsh environment in the Reactor Building for the accident they are required to mitigate, during the time they are required to function. Any unqualified components in the motor operators will be replaced with qualified components by November 1985.

i' JCO-0C-84-5, Rev. O December 19, 1984 Page 1 of 2

0YSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) b 4y.

8

rg a.

N

L,/"

a-JCO-0C-84-5, Rev.'O

-December 19. 1984 9-

/

Page 2 of 2-Y'.

[,

COMPONENTS Tag Numbers ~

RE-04A,'RE-048, RE-04C, RE-04D:

7 Description, f51gh Drywell; Pressure Switches-

.0BJECTIVE' h objective offthis discussion is.to determine that there is a basis for iconcluding that' the existing component' will perform its required function.

s

. COMPONENT LOCATION

-All of these pressure switches are located in the northwest-quadrant of-elevation'51'3" in the Reactor Building.

COMPONENT FUNCTION These. switches' generate scram, drywell and reactor building isolation signals

and = initiate the Standby Gas. Treatment.

s3 EVALUATION

-Mbese switches are' located outside the primary containment and are required to mitigate events inside the containment. The environment in the Reactor Building is not' expected to become harsh due to a break inside the1 containment, and hence these components are expected to function. The switches may be'in a s'

harsh environment for breaks outside containments however, they are not required to function to mitigate these events. Their failure for outside containment breaks will not mislead the operator since he will know that drywell temperature is normal and that drywell sump level is normal.

The operator may initiate containment spray if he thinks that he has exceeded the "drywell pressure action level in the E0Ps. This, however, should not result in unacceptable. consequences.

CONCLUSION

.4 The switches will perform their function whenever they are required to perform their safety function.

o GUALIFICATION PLAN An evaluation will be performed to confirm that these components are not

. exposed to a harsh environment in the Reactor Building for the accident they are' required to mitigate, during the time they are required to function. Any qualified switches will be replaced with' qualified switches during the

~

November 1985 outage.

l' l'

JCO-0C-84-6, Rev. 0 December 19, 1984 Page'l of 2 0 -

3 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) s

t, 4

- /]

y}

'o t

A L: -

xy

- p

[

JCO-0C-84-6, Rev. 0-

'i December 19, 1984.

M'

.Page 2 of.2 h

COMPONFNTS '

nd,

, j Tag; Numbers 1ID-45; ID-46A,.ID-46B

' Description-LID-45.is RPV Narrow Range Pressure Transmitter, ID46A,B are RPV4 m.

Wide Range Pressure Transmitters..

OBJECTit'E -' u w

The objective;of this' discussion is-to' determine that the failure of these

. components will'not. degrade other safety functions or mislead the operator.

.a

,~

COMPONENT LOCATION

'E These-transmitters are located.in the Reactor Building on elevation 51'3".

JID-45 and ID-46B-are located in the southeast. quadrant c.nd-ID-46A is located

=in'the northwest' quadrant.

6 in

?.

COMPONENT FUNCTION These-transmitters feeo the wide range and the narrow range indicating

~ instruments in the control room.' They do not provide any safety related E

function.

EVALL'ATION

>These transmitters are located outside the primary containment. For breaks inside the containment, they will function normally. For breaks outside the containment, their; failure will not affect the operation of any safety system..RPV' pressure indication may be known from the status of the' safety valves or core spray. As.long as. level-indication is-available to the operator, on YARWAYS, no safety function is violated.

If level is undetermined,.then the operator has'to depressurize manually before flooding the vessel using core spray. RPV prevsurel indication is not. required since.

~ core spray injection is automatic once RPV ! pressure drops below 285 psi..

Also,'these transmitters are located in different areas of the Reactor Building such that they will.not see a harsh environment simultaneously.

sg

'Their.. failure will not mislead the operator since no operator. action is reguired by procedures if RPV pressure canriot be determined for breaks outside

- A -containment.

' CONCLUSION 3;

[

M. These transmitters'do not. provide any, safety related function with respect to N

the auto actuation of a safety system. A singic break should not result in the failure of all of these components such that pressure indication would not W

D-

.::!:be completely lost.

s d QUALIFICATION PLAN The transmitters will be replaced with qualified transmitters during the 1

7

. November 1985 outage.

n.

i

JCO-0C-84-7, Rev.10 December 19, 1984 Page 1 of 2-4 i

OYSTER CREEK NUCLEAR GENERATING STATION

- JUSTIFICATION FOR CONTINUED OPERATION (JCO)

~

e l

l A

L l

[:

l

=-

JCO-OC-84-7, Rev. O December 19, 1984 Page 2 of 2 COMPONENTS

,o~

Tag Numbers IG-06A, IG-06B Description.

Isolation Condenser Secondary Side Level Transmitter OBJECTIVE

- The objective of this discussion is to determine that the failure of the identified components as a result of a harsh environment will not degrade other safety functions or mislead the operator.

~

COMPONENT LOCATION

' These transmitters are located in the east area of the Reactor Building on the-95' elevation.

- COMPONENT FUNCTION

- The components sense and transmit to the control room the water level on the secondary (shell)-side of the isolation condensers.

EVALUATION The isolation condensers can operate for at least 45 minutes each without makeup tofthe secondary side. Makeup water is added from the Condensate

' Transfer System by operator action. Loss of level indication would not prevent the_ operator from adding makeup to the shell side of the isolation condensers. Since this is normally a manual action and part of normal training, the operator would makeup to the secondary side in less than 45 minutes with or without level indication.

' If, after ed 45 minutes (or 100 minutes if 2 condensers are in service), the operator does not take action to makeup to the secondary side, heat transfer capacity will be reduced, and the reactor vessel will begin to re-pressurize.

The operator has the means to determine pressure and can take action to depressurize the RPV by either blowing down the vessel or refilling the isolation condenser shells from the Condensate Transfer System. Status of the condensate transfer pumps is available in the control room.

CONCLUSION I-The failure of the level transmitters will not degrade the function of the isolation coadensers since the system can operate for up to 100 minutes on the inventory of water contained on the shell side of both condensers. The failure of these transmitters will not mislead the operator since this is normally done manually and part of operator training.

QUALIFICATION PLAN These transmitters will be replaced with qualified transmitters during the i

November 1985 outage.

l

~, _ _ _. -..

c.

JC0-0C-84-8, Rev. 0 December 19, 1984 Page 1 of 2 r

OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)

I l-( -

l l.

i

-m

m=

JCO-0C-84-8, Rsv. 0 December 19, 1984 Page 2 of 2

COMPONENTS-

,e' Tag Numbers-

.ID-13A,~ ID-13B

~ Description iGEMAC Water Level Transmitters OBJECTIVE The objective of this discussion'is to determine:

that the failure of -these components will not degrade other safety functions nor mislead the operator; that the function may be accomplished by some other qualified

• ~

components.

. COMPONENT LOCATION Transmitter ID-13A-is located in the northwest quadrant of the Reactor-Building on'the 51'3" elevation. Transmitter ID-13B is located in the southeast ~ quadrant of ~ the Reactor Building on the 51'3" elevation.

' COMPONENT FUNCTION These level transmitters are only one of several means used by the operator

.for indications of RPV water level..These transmitters have nearly the same span as the YARWAY transmitters and.thus provide redundancy to the YARWAYS.

One of two GEMACs is selected to provide an input to feedwater level control which is not safety related.

EVALUATION

--These_ transmitters.are located in the Reactor Building and thus would not be in a harsh environment for breaks inside the drywell. For breaks outside the the loss of these transmitters does not prevent any plant safety containment,

. function from occurring since no safety-related trips are based on the GEMAC transmitters. The two GEMAC transmitters are located in different areas of the Reactor. Building and will not see the same harsh environment simultaneously.

The GEMAC transmitters provide adverse level indication to the YARWAY level

~

instruments which are the operator's primary level indication. Also, the operator would use additional control room indication to make his Even.in the event the operator judges that he cannot determine determination.

the level, the E0Ps provide the~ operator with guidance to prevent core uncovery.

CONCLUSION The failure'of these transmitters does not significantly degrade the ability of the operator to monitor the RPV water level. There are no safety related trips which are based on the GEMAC transmitters.

QUALIFICATION PLAN

-These' transmitters will be replaced with qualified transmitters during the

. November 1985 outage.

b JCO-0C-84-9, Rev. O December. 19,'.1984

- Fage 1 of 2.

P OYSTER-CREEK NUCLEAR GENERATING STATION.

- JUSTIFICATION FOR CONTINUED OPERATION (JCO)

(

2 h

I F.

s Y

r i

u 1

3 :

9;~

}

JCO-0C-84-9, Rsv. O December 19, 1984

+

Page 2 of 2

,. COMPONENTS

,o~

-n

' Tag. Numbers l-RV-29A, RV-295,-RV-29C, RV-29D'~

Description These are pressure' switches which start core spray booster pumps. based on core spray main pumps discharge pressure.

OBJECTIVE O

The objective of this' discussion isito show that the failure of these switches

'will.not degrade _ core spray safety function and will not mislead the operator.

~

COMPONENT LOCATION These switches are located in the Reactor Building on the -19'6" elevation.

.RV-29A.C_are located in the northwest corner room; RV-29B.D are located in the southwest corner room.

4 COMPONENT FUNCTION L

These pressure switches start core spray booster pumps based on dore spray

~

main' pumps discharge pressure.

EVALUATION s

These switches are located outside the containment, hence for breaks inside the containment, they will perform their safety function as required.

If they fail during breaks'outside the containment, the operator can manually start the booster pumps from the control room _after the main pumps have been started. Main pumps and boostee pumps light indicators are available in the control room.

CONCLUSION The operator can take manual control, hence the failure of these switches will not degrade core' spray safety function nor mislead the operator.

QUALIFICATION PLAN These. switches will be replaced with qualified switches during the November 1985 outage.

{

g

.s JCO-0C-84-10,. Rev. O December 19, 1984.

Page 1 of 2 4

9 OYSTER CREEK NUCLEAR GENERATING STATION.

JUSTIFICATION FOR CONTINUED OPERATION (JCO) n l

I

JCO-0C-84-10, Rev. O December 19, 1984 Page 2 of 2

-COMPONENTS

.e~

Tag' Numbers' Description'

~ RV-40A, RV-40B, RV-40C, RV-40D These are pressure switches which close on core spray booster pump discharge pressure and inhibit the start of the backup booster pump.

OBJECTIVE

.The objective'of this discussion is to_ determine that the_ failure of these switches will not degrade core spray safety function and will not mislead the operator.

COMPONENT LOCATION

The RV-40A C switches are located in the. northwest quadrant in the Reactor Building on the 51'3" elevation. The RV-40B,D switches are located in the' southwest quadrant in the Reactor Building on the 23'6" elevation.

COMPONENT FUNCTION These pressure switches inhibit the start of the backup booster pump once the-main booster pump starts.- They close when booster pump discharge pressure reaches the switch setpoint.

! EVALUATION

-These-switches are located outside the containment, hence for breaks inside the containment, they will perform their safety function as required. For breaks outside the containment, if these switches fail and the backup booster-pump starts in addition to~the main booster pump, the operator can manually

trip the backup booster from the control room. All core spray pumps (including the backup _ booster. pumps) have light indicators in the control

room.' Further, the_RV-40A.C and the RV-40B,D switches are located in different areas of the Reactor' Building and will not simultaneously see the harsh environment.

CONCLUSION The operator. can take manual control and has control room indication, hence

-the failure of these switches will not degrade core spray safety function nor mislead the operator.

. QUALIFICATION PLAN These switches will be replaced with qualified switches during the November 1985 outage.

i l

l

m C

JCO-0C-84-11, Rev. 0

. L ~

. December 19, 1984

- Page 1 of 2 l

+

t OYSTER CREEK NUCLEAR GENERATING STATION hSTIFICATIONFORCONTINUEDOPERATION(JCO)

['

JCO-0C-84-11, Rev. O December 19, 1984 Page 2 of 2 COMPONENTS Tag Numbers

'RV-26A, RV-26B-Description

. Core: Spray System Flow Transmitters

~0BJECTIVE The-objective of this discussion is to determine:

'that the. failure of these components will not degrade other safety functions nor mislead ~ the operator; that the function may,be accomplished by some other qualified components.

COMPONENT-LOCATION The RV-26A transmitter is located in the northwest _ quadrant of the Reactor

' Building on elevation 51'3".

The RV-26B transmitter is located in the southwest quadrant of the Reactor Building on elevation 75'3".

COMPONENT FUNCTION These' flow transmitters. provide core spray flow indication to the operator fduring a break condition.

EVALUATION

The environment in the Reactor Building is not expected to become harsh due to

~

a break.inside the containment, and hence these components are expected to

' function.

For breaks outside the drywell, the operator may confirm Core Spray System operation using RPV water. level in the event that the flow transmittersr were lost due to a harsh environment.

In addition, the. operator has

. indication that the core spray pumps are running and that the valves are open. The flow transmitters provide no input to other safety systems and thus will not impact other safety functions. Further, the transmitters are located in different areas of the Reactor Building and will not simultaneously experience the harsh environment.

CONCLUSION The failure of these transmitters does not preclude the operator from confirming Core Spray System operation and does not affect other safety functions.

QUALIFICATION PLAN These transmitters will be replaced with qualified transmitters during the November 1985 outage.

1 JC0-0C-84-12,-Rev. 0 s

December 19, 1984 Page 1 of 3 9

f 4

OYSTER CREEK NUCLEAR GENERATING STATION f

JUSTIFICATION FOR CONTINUED OPERATION {JCO) s v

.a 3

r

(-

l

m JCO-0C-84-12, Rsv. O December 19, 1984 Page 2 of 3 COMPONENTS Tag Numbers

'IP-03A, IP-03B; IP-05A, IP-05B, IP-05C, IP-05D*,'V-3-87, V-3-88 Description Containment Spray Flow Transmitters (IP-03A B)

' Containment Spray Heat Exchanger dP Transmitters (IP-05A,B,C,D)

Containment Spray Heat Exchanger ESW Outlet Valve Motor Operators (V-3-87,88) x OBJECTIVE The objective of this discussion is'to determine:

that the' failure.of'these components will not degrade other safety functions nor mislead the operator; that there is a basis for concluding that the existing components will perform their required function.

COMPONENT LOCATION Transmitters IP-03A, IP-05A,B and valve V-3-87 are located in the northeast quadrant of the Reactor Building on elevation 23'6".

Transmitters IP-03B, L

IP-05C,D and valve V-3-88 are located in the southeast quadrant in the Reactor l

_ Building on elevation 23'6".

COMPONENT FUNCTION These components' provide pumpiag capability and pwnp start logic' for the Containment Spray System as well as operator indications to confirm its performance. The Containment Spray System is required to remove heat and reduce pressure inside the containment for breaks inside the drywell.

EVALUATION These components are required to function for breaks inside the drywell.

However, the components are located outside containment. The environment in the Reactor Building is not expected to become harsh due to a break inside the containment, and hence these components are expected to function. For breaks outside the drywell, the Containment Spray System is not required to mitigate l

the event. In addition, the failure of the instrument components will not mislead the operator because he can determine from drywell parameters that the j.

l break is not inside containment aad that for outside containment breaks, the L

Containment Spray System is not required.

If containment spray should start

- inadvertently as a result of the failure, there should be no safety Further, there is control room indication of pump actuation.

consequence.

t I

CONCLUSION The failure of these components does not degrade the effectiveness of the Containment Spray System to perform its function because the components will l

function for events for which they are required to operate.

~ ~

p JCO-0C-84-12, Rev. O December 19, 1984 Page 3 of 3

-QNALIFICATION PLAN

.o~

In evaluation will be performed to' confirm that these components are not exposed to a harsh environment in the_ Reactor Building for the accident they are required to mitigate, during the time they are required to function. Any.

unqualified _ transmitters will be replaced.with qualified transmitters during the November 1985 outage. The unqualified components in the motor operators will be replaced with qualified components during the November 1985 outage.

i e

I i

E.

V JCO-0C-84-13,.Rev. 0

.g%

-December 19, 1984.

Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 1

.4

/

I m...

^

f JCO-0C-84-13, Rev. O December 19, 1984 Page 2 of 2

-COMPONENTS ~

je

, Tag Number IP-07 Description-

.Drywell' Pressure Transmitter OBJECTIVE The objective of this. discussion is to determine:

that the failure of these components will not' degrade other safety functions nor mislead the operator to take an' unsafe action; there is a basis for concluding that the existing component will perform its required function.

COMPONENT LOCATION This transmitter is located in the northwest quadrant in the Reactor Building on elevation ~51'3".

COMPONENT FUNCTION This' transmitter provides control room indication of drywell pressure-for breaks inside containment.

. EVALUATION.

This component is required to function for breaks inside the drywell.

The environment in the However, the component is' located outside containment.

Reactor Building is not expected to become harsh due to a break inside the The failure containment, and hence these components are' expected to function.

of the instrument will not mislead the operator because he has drywell temperature and sump level indication.

If he thinks that drywell pressure has reached an E0Ps action point, however, he would initiate containment spray.

This should have no adverse consequences.

CONCLUSION The transmitter is expected to be in a mild environment for those events for In addition, the failure of the transmitter will which it is to be operable.

not mislead the operator.

QUALIFICATION PLAN This transmitter will be removed, and its function will be performed by IP-15 (a_ qualified transmitter) during the November 1985 outage.

'l

+

.L i fI

> lp,;-

JCO-0C-84-14, Rev. 0

. December 19, 1984 Page 1 of 2 e

i OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) b.

L 4

3. 4 4

if s

-O 2

- 4,.

s 4

x a

JCO-0C-84-14, Rev. O December,19, 1984

.Page 2 of 2 COMPONENTS ~

. ~. ' "

1 Tag Numbers:

~ RE-02B; RE-02C, RE-02D

. Description LOW-LOW Level Switches OBJECTIVE. '

The' objective of this discussion is to determine'that the failure of these components will.not degrade other safety functions 'or mislead the operator.

COMPONENT LOCATION-RE-02A.C are located on rack RK01 in the northwest quadrant of the Reactor

. Building at elevation 51'3".

RE-02B,D are' located on rack RK02 in the southeast quadrant of;the Reactor Building at ele'istion 51'3".

' COMPONENT FUNCTION The. Low-Low switches function is to turn on core spray pumps, containment

= spray (pumps (high drywell pressure is also needed), primary and secondary containment isolation, trip-recirculation pumps, and initiate the isolation condensers..

EVALUATION-If-the break is inside.the containment, the switches will perform their safety function as required. For any condition which results in Low-Low level, the operator is trained to' initiate msnually all systems which should have

' initiated automatically but failed to do so.

It is a general operator training concept that the operator is.to backup all automatic actuations..There are sufficient control room indications to do sc.

This is not a misleading condition. For large and intermediate breaks, it is expected that Low-Low level will befreached very quickly before the effect of the harsh environment

. becomes severe. Further, pairs of switches are located at different locations in the Reactor Building.and would not be susceptible to the same harsh

. environment simultaneously. For a small break, even though Low-Low may not be i

reached quickly, the operator would have a longer period of time to act, and the break would not affect more than one pair of instruments so that auto actuation would occur.

+

- CONCLUSION The operator should manually perform all actions required by Low-Low setpoint j

' in-caseithe' switches fail. Both pairs of switches would not be subjected to the same harsh environment because of their diverse locations.

e

~

I QUALIFICATION PLAN j

I' These components will be replaced with qualified components during the

~

. November 1985 outage.

I 5

' JCO-0C-84-15, Rev. O December 19, 1984 -

Page 1 of 2' 0YSTER CREEK NUCLEAR GENERATING STATION t

JUSTIFICATION FOR CONTINUED OPERATION (JCO)

JCO-0C-84-15, Rev.-0 December 19, 1984 Page 2 of 2 COMPONENTS' f

' Tag' Numbers DPS-66A,.DPS-66B Description Switches to open vacuum breakers based on Reactor Building to torus dP

.0BJECTIVE.

.s-The objective of this discussion is.to determine that the failure of these components will not; degrade _other' safety functions or mislead the operator.

COMPONENT LOCATION These components are located in the southeast quadrant of the Reactor Building

- at= elevation 23'6".

COMPONENT FUNCTION These switches provide opening and closing signals to the torus - Reactor

~

These vacuum breakers are used to purge air from

' Building vacuum breakers.

the Reactor Building into the torus to prevent exceeding the negative design pressure;of the containment during an event.

EVALUATION The environment in the Reactor Building is not expected to become harsh due to a break inside the containment, and hence these components are expected to f un'etion. For breaks outside the containment, a harsh environment may cause these' switches to fail.. However, for breaks outside containment, it is not expected that there will be a need to open these vacuum breakers since containment ~ spray would not normally be required.

If these switches fail and they are needed, the operator can manually open.and close the vacuum breakers from the control room. The failure of these switches will not mislead the

. operator since there are indications of vacuum breaker position in the control room.

CONCLUSION The failure of these switches will not prevent the operator from manually

. opening and closing the torus to Reactor Building vacuum breakers from the control room. The operator has indication of vacuum breaker position available so that he will not be misled.

QUALIFICATION PLAN These components will be replaced with qualified components during the November 1985 outage.

B e

-n.

_,g.

{

~

f

(-

a t!

JCO-OC-84-16, Rev. 0 December 19, 1984,-

Page 1 of 2 i

't P

T OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)

- -- ~_.

.JCO-0C-84-16, Rav. O

?

December 19. 1984 Page 2 of 2 g-COMPONENTS

.4 7- '

l Tag. Number:

- VMS (Valve MonitoringLSystem)

Description Monitors status of EMRVs and SVs OBJECTIVE The objective of:this discussion is to determine that1the failure of these

' components will. not degrade other safety functions nor mislead the operator.

COMPONENT LOCATION The acoustic monitors for.the VMS are located inside the drywell.

COMPONENT FUNCTION The VMS is used to detect flow through or leaks from an EMRV or SV.

The sensing components are all located inside the drywell and h'ence are unaffected by a break outside the containment.

EVALUATION For breaks-inside containment, the operator can determine f rom other parameters (tail pipe temperatures, EMRV position indication) that an EMRV or SV is open in the event that the VMS fails.- Following a LOCA, it is not likely that a SV'will be required to function, therefore, there is no concern for_a stuck SV.

For a'small break LOCA, ADS will actuate and cpen all EMRVs.

A stuck open EMRV will'not be a concern under these conditions. For a large break LOCA.~ the system will be completely depressurized without EMRV actuationc~Thus,1the loss of the VMS has no safety significance for a break l

Jinside containment. For a break outside containment, the VMS is unaffected and will function normally. The failure of the VMS does not affect the normal operation of the EMRVs or the ADS.

CONCLUSION l

The failure of the VMS will not degrade any other plant safety function nor mislead the operator.

-QUALIFICATION PLAN GPUN is currently performing an engineering evaluation to' determine how the

-system can be modified to be environmentally qualified.

5

JCO-0C-84-17,.Rev. O December 19, 1984,,

Page 1 of 2 9

OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) o

e JCO-0C-84-17. Rev. O December 19, 1984-Page 2 of 2' COMPONENTS.

p

-Tag Numbers

V-20-3, -- V-20-4, V-20-33, v-20-32

} Description' Core Spray _ Pump Suction Isolation Valve Motor Operators OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions nor mislead the operator.

' COMPONENT LOCATION

- V-20-3 and V-20-33 are located in the northwest corner room of the Reactor Building at elevation 29'6".

V-20-4 and V-20-32 are located in the southwest corner room at elevation 19'6".

COMPONENT FUNCTION These valves are normally open and are closed by the operator to isolate's' core spray line break in a-corner room.

EVALUATION For breaks inside or outside the containment, these valvds are not required.to change position in order to mitigate the event. Thus, the failure of the Also, the control room operator has valve operator will not affect the event.

indication of the position of the valves as well as manual switches in order to confirm that they are open.

' CONCLUSION-These valves are normally open and are required to remain open in order to The operator has position indication for the valves as mitigate an event.

well as manual control switches to confirm that the valves remain open.

QUALIFICATION PLAN The unqualified components in the motor operators will be replaced with qualified components during the November 1985 outage.

JCO-0C-84-18, Rev. 0 December 19, 1984 Page 1 of 2

)

OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATIONFORCONTINUEDOPERATION(JCO) i i

)

JCO-0C-84-18, Rev. O December 19, 1984 Page 2'of 2 COMPONENTS ~

Tag Numbers V-20-12. V-20-18, V-20-26, V-20-27 Description-Core Spray Pump Test Isolation Valve Motor Operators (V-20-12, 18)

Core Spr'ay System Test Valve Motor Operates (V-20-26, 27)

OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions nor mislead the operator.

COMPONENT LOCATION.

V-20-12 and V-20-27 are located in the northwest quadrant of the Reactor Building at elevation 51'3".

V-20-18 and V-20-26 are located in the southwest quadrant of the Reactor Building at elevation 75'3".

COMPONENT FUNCTION These valves are.used for the monthly surveillance testing of the Core Spray System to recirculate water through the pumps and back to the torus.

EVALUATION The environment in the Reactor Building'is not expected to become harsh due to a break inside the containment, and hence these components are expected to

,5-function. The components are in their desired post-accident positions except during the monthly surveillance testing. During this test, only one subsystem is tested at any one time. The operable subsystem provides sufficient capability to satisfy the function of the Core Spray System. The operator would not be misled since he has valve position indication available in the

. Control Room.

CONCLUSION The failure of these valves would not degrade the Core Spray System since one subsystem will always be available for operation to mitigate an event. The operator will not be misled since valve position indication is available in the Control Room.

QUALIFICATION PLAN n

An evaluation will be performed to confirm that these components are not exposed to a harsh environment in the Reactor Building for the accident they are required to mitigate, during the time they are required to function. Any unqualified components in the motor operators will be replaced with qualified components during the November 1985 outage.

1 l

JCO-0C-84-19, Rev. 0 December 19, 1984 Page 1.of 3_

W h

OYSTER CREEK NUCLEAR GENERATING STATION

+.

JUSTIFICATIONFORCONTINUEDOPERATION(JCO) ev.-n

-,-,--n---

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

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

D

/

m

.JCO-0C-84-19. Rev.~0 December'19, 1984-Page 2 of 3

./0 COMPONENTS Tag Numbers' _ IP-18A, IP-18B Lescription LContainment Spray-Pump Hi Temperature Trip OBJECTIVE L

The objective of this discussion is to determine:

that there is's basis for' concluding that the components will perform their intended function;

+

4

-that fail.ure of the~ components will not degrade other safety systems

=*

or mislead the operator.

J COMPONENT LOCATION i.

These switches are located on the discharge piping of the containment spray

-heat exchangers at' elevation 23'6" in the Reactor Building.

IP-18A is in the northeast-corner and IP-18B is in the southeast corner.

COMPONENT FUNCTION (These components function to trip the containment spray. pumps upon receipt of a high temperature indication at the outlet of the. containment spray heat exchangers.

' EVALUATION These components areflocated outside'the drywell on the containment spray heat exchangers discharge piping. The Containment Spray System is only required to

-function to mitigate breaks inside the drywell.- The environment in the

~

Reactor Building is not expected to become harsh due to a break inside the.

containment, and hence these components'are expected to function.

In the

(

event of a-break outside containment, the Containment Spray System is not required to operate, and.the failure of the temperature switches due to the

.resulting environment.will not cause an unsafe condition.

These switches do not interface with or control any other components and their failure would not affect any other safety or accident mitigation systems.

-The: failure of these switches will not mislead the operator since there exist

~

separate temperature elements for each system which will give indication of heat exchanger discharge temperature in the control room.

CONCLUSION

-These' components will function during an inside containment LOCA when the

Containment-Spray System is required to operate. The failure of the

. components due to a harsh environment created by an outside containment break does not degrade any other safety systems nor will it mislead the operator.

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

N>

JCO-0C-84-19,'Rev. O December 19, 1984 Page 3 of 3 C"ALIFICATION PLAN An evaluation will'be performed to confirm that these components are not exposed to a harsh environment in the Reactor Building for the accident they Any are: required to mitigate, during the time they are required to function.-

alified components.will.be replaced with qualified components during the unqu 7,,

~

- November 1985 outage.

.. l

\\

'I 3 y 1 J.

?

~

m.

I

.,. r,; -

\\

g JCO-0C-84-20, Rev. O December.19,-1984

Page 1 of 4 O

m

-OYSTER. CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)

O o

JCO-0C-84-20, R2v. O

-*1 December 19, 1984 Page 2 of 4

. COMPONENTS.

. o..~ ~ -

LCable:-

Rockbestos Firewall and Rockbestos EP.

DESCRIPTION-

-Manufacturer:.5The Rockbestos Company Firewall EP Firewall III Model: 1 Function:

Power Cable ~

Control Cable

-Voltage:

600' Volts 600 Volts-90*C Rating:

.90*C -

Cross Linked Polyethylene Insulation:

-EPR.

Jacket:

Hypalon Hypalon 0BJECTIVE The' objective of this engineering justification for continued operation'is to demonstrate that.(1) the Rockbestos cable will perform its safety function in the event ~of a design basis accident at DCNGS, and.(2) that the plant can be safely operated in the interim until the completion of the environmental qualification program by,the Rockbestos Company.

EQUIPMENT FUNCTION The cable connects nuclear safety-related equipment in the plant. The most A

-severe function with respect to loading and environment has been evaluated.

combination of analysis and inadequately documented tests indicates that cable

.is acceptable for the intended functions.

EQUIPMENT LOCATION The cable is installed in the Reactor Building, including the drywell. High radiation areas which include the cleanup demineralizer room and cleanup room

'do not contain any nuclear safety related equipment.

EVALUATION An' evaluation of current test information and analyses of the Rockbestos cable indicates that full qualification is not demonstrated; however.. partial test data does provide a basis for concluding the cable will perform its function.

These findings are based upon the results of both the NRC audit of Rockbestos (Reference 6) and the GPUN audit accomplished August 21-23, 1984 (Reference 1).

GPUN has-concluded, as has the NRC staff (Reference 3), that "at this time no immediate safety problem exists in the use of Rockbestos cables".

JCO-0C-84-20, Rev. O December 19, 1984 Page 3 of 4 ItwastheintentofRockbestostoconducttestsinaccordance,[ithIEEE Standards 323-1974 and 383-1974.

Inadequate traceability, inidequate documentation and the general poor auditability of the supporting documentation does'not conform with NRC requirements. The responses on the part of Rockbestos Rockbestos tests provides limited support, however, for partial test data.

and analyses on Firewall EP cable (Reference 5) demonstrated a thermal qualified life of 40 years at.90*C and radiation tolerance of 2 x 10' rads.

Franklin Research Center tests (Reference 10) on Firewall III with a Neoprene l i Since

. jacket instead of Hypalon, indicated acceptability of the insu at on.

Hypalon is superior to Neoprene for thermal qualified life and in light of the Rockbestos. test results, it is anticipated that the Firewall III cable for nuclear service is superior to that with Neoprene.

As a consequence of generic problems with the Rockbestos test program for these Class 1E cables (Reference 3), Rockbestos has committed to conduct a new This will be completed in July 1986.

In the meantime, CPUN test program.

Also GPUN will verify the results of the Rockbestos supplemental program.

The periodic conducts its surveillance on nuclear safety related equipment.

review of cable performance will provide a measure of confidence of the performance function of the cabla; any indications of degradation will be evaluated for its potential impact on the cable performance during and after the accident.

EQUIPMENT QUALIFICATION PLAN GPUN will evaluate the results of the Rockbestos test program to assure it is Upon completion of this responsive to the CPUN and NRC audit findings.

program, the EQ files required in accordance with 10CFR50.49 will be completed.

In the meantime, the results of GPUN plant surveillance, which includes safety equipment utilizing this cable will be evaluated and the impact on cable A trend analysis of IR values for these power cables performance assessed.

has already been initiated.

CONCLUSION GPUN concludes that the OCNGS can be safely operated pending completion of equipment qualification as required by 10CFR50.49, Section i, and the NRC This letter of May'25, 1984 " Request for Additional Information".

consideration includes, as appropriate,. items 1 through 5 as follows:

Item 1 - Accomplishing the safety function by some designated alternative equipment if the principal equipment has not been demonstrated to be fully qualified.

This is not appropriate for the equipment involved.

The validity of partial test data in support of the original Item 2 -

qualification.

r:

1

~

j l

1 JCO-0C-84-20, Rev. 0 l

December 19, 1984 Page 4 of 4 This is the basis for justification for continued operation greprovided above.

Item 3 Limited use of administrative controls over equipment ~that has not'been demonstrated to be fully qualified.

This is not appropriate for the equipment involved.

Item 4 - Completion of the safety function prior to exposure to the accident environment resulting from a design basis event and ensuring that the subsequent failure of the equipment does not degrade any safety function or mislead the operator.

-This is not appropriate for'the equipment involved.

Item 5 - No significant degradation of any safety function or misleading information to the operator as a result of failure of equipment under the accident environment resulting from a design basis event.

This is not' appropriate for the equipment involved.

Based upon the evaluation provided in Section 6, GPUN concludes that the cable is qualified to perform its safety function in the interim period before completion of the Rockbestos tests. No significant degradation af any safety function or misleading information to the operator is expected under the accident environment resulting from a design basis event.

REFERENCES 1.

GPUN Memorandum QA-D/P-84-828 dated November 30, 1984 (Finding No. 3).

Oyster Creek Nuclear Generating Station - Environmental Qualification of 2.

Safety Related Equipment dated November 1, 1980.

3.

IE Information Notice No. 64-44 dated June 8, 1984.

4.

File EQ-0C-311, Revision 0, Rockbestos EP.

5.

~Rockbestos Report No. QR 1804 dated April 6, 1981.

6.

NRC Trip Report - Audit of Rockbestos Company Qualification Documents, dated September 12, 1983.

7.

GPUN - TDR 297, Revision 1 dated August 19, 1982.

8.

Letter to Kearny (GPUN) f rom Littlehales (Rockbestos) dated June 28, 1984.

9.

File EQ-0C-312, Revision 0, Rockbestos Firewall.

10. ' Franklin Research Center Final Report F-C3798 dated March 1974.

1 JC0-0C-84-21, Rev. 0 December 19, 1984,,

Page 1 of 2 D

-e

' OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 4 i

i a*

JCO-0C-84-21. Rzv. O December 19, 1984 Page 2 of 2 COMPONENTS

1. 7-Tag Numbers DC-2 Description 125V DC Power Supply OBJECTIVE

'The objective of'this discussion is to determine that fa. lure of this

-component will not degrade other safety systems nor misitad the operator.

COMPONENT LOCATION

'This power supply is located in the northeast quadrant of the Reactor Building on elevation 75'3".

COMPONENT FUNCTION This component supplies power to one'of the isolation condenser condensate

. valves'(V-14-35).

EVALUATION -

The component sees a harsh environment for the isolation condenser break outside the drywell.

It is.likely that DC-2 satisfies its function before a harsh environment causes it.to fail.

In this case, the loss of DC-2 will not prevent the operation of the -intact isolation condenser or adversely affect any other. safety function.

'If DC-2 failed prior to satisfying its function, decay heat removal would be This mode through the EMRVs with RPV inventory makeup supplied by Core Spray.

could be sustained for a. lengthy period due to the large heat capacity of the torus pool until the Reactor Building could be entered, and the closed isolation condenser valve could be manually opened.

The operator-would not be misled by the loss of DC-2 since the valve position I

indicating lights for V-14-35 would both go out signifying that the power to the valve has been lost.

CONCLUSION

-The loss of.DC-2 will not degrade other plant safety functions and will not provide misleading information to the operator.

QUALIFICATION PLAN DC-2 will be qualified or replaced with a qualified substitute during the November 1985 outage.

i

)

i JCO-0C-84-22, Rev. O December 19, 1984e-Page 1 of 2 W

OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) t I

.7

s

-JCO-0C-84-22. Rev.LO

-December 19, 1984

Page 2 of 2

=

.CetPONENTS

. Tag Numbers

,.V-24-30 Solenoid valve V-6-917-Description i Drywell Isolation Valve Solenoid Operator-OBJECTIVE ~

-The objective of this discussion is to determine:

that theLfailure of this component vill not degrade other safety functions nor mislea( the operator; that the safety function can be accomplished by some other qualified

. equipment.

COMPONENT LOCATION This equipment is located in the southwest quadrant of the Reactor Building at elevation ~51'3".

COMPONENT FUNCTION

-This equipment functions to isolate'the drywell upon a containment isolation signal.

EVALUATION 1he' environment in the Reactor Building is not expected to become harsh due to a _ break iriside the containment, and hence these components are expected to

' function..For breaks outside containment, this equipment is in series with a redundant qualified isolation valve (V-24-29) which is located inside 1 containment and would function to isolate the drywell upon a receipt of a containment isolation signal.-

Both the subject valve a.'.d the redundant qualified valve have posit'nn indication given in the contrcl~ room, thus a' failure of the valve w:oid not mislead the operator.

. CONCLUSION.

The failure of this equipment.will not degrade the isolation of containment since it is.in series with a qualified valve which will perform the intended function. The operator would not be misled by failure of this component.

QUALIFICATION PLAN The unqualified equipment will be replaced with qualified components during the November 1985 outage.

JC0-0C-84-23 Rev. O December 19, 1984 Page 1 of 2-

.0YSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) k f

4-r

JCO-0C-84-23, Rev. 0 December 19, 1984 Page 2 of 2

. COMPONENTS-p-

~

~

Tag. Numbers RE-05A, RE-05B; RE-05/19A, RE-05/198 Description RPV Level Switches OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions or mislead the operator.

COMPONENT LOCATION

.RE-05A and RE-05/19A are located on rack RK01.in the northwest quadrant of the

-Reactor Building:at: elevation 51'3".

RE-05B and RE-05/198 are located on rack RK02.in the southeast quadrant of the Reactor Building at elevation 51'3".

COMPONENT FUNCTION-The function of the level switches is to generate a scram signal on low RPV level and a turbine trip on high RPV level. Further, RE-05/19A,8 provide

YARWAY. level indication in the control room.

. EVALUATION For breaks inside the containment these switches will perform their safety

. function as. required.. For large breaks outside containment, a low level scram would be expected to occur very quickly. Redundancy exists via a low pressure MSIV closure scram or. turbine trip on low steam flow. Also, a Low-Low level MSIV closure scram would also occur if low level scram failed. For small breaks outside containment,- the redunda.it scrams will be sufficient to assure that the-scram function is accomplished. Further, the switches are located at different locations in the Reactor Building such that they will not see the same harsh conditions simultaneously. The operator has diverse level instrumentation in the control room so that the loss of one YARWAY indicator or even both would still not result in a' misleading condition. However, only one indicator would be expected to fail from a single break location.

2 CONCLUSION.

The failure of these switches will not degrade the scram function which would

~ occur by diverse means, or they may perform their scram function before the harsh environments occur.

In addition, the operator can scram manually.

There.is sufficient diversity of level instrumentation to prevent a misleading condition. Further, only one of the two YARWAY indicators would be expected to. fall from a single break location.

QUALIFICATION PLAN These switches will be replaced with qualified switches during the November 1985 outage.

l

g.

s.

ATTACHMENT II COMPONENTS REPLACED WITH QUALIFIED COMP 0NENTS

.s-IN PREVIOUS OUTAGES Plant ID#

Component System ll)

TB #63-242 Terminal Block MSIV (NS03A) 2)-

.TB #63-246 MSIV (NS038).

-3)

TB.#22-389 MSIV By-Pass (V-1-106) 4)

TB #22-390 MSIV By-Pass (V-1-107) 5)

'TB #71-412_

EMRV's 6)

TB #22-640 Drywell Floor Drain Sump 7)

TB #22-641 Drywell Equipment Drain

8)-

V-38-22 Solenoid Valve Torus Sample 9)

V-38-23 10)

V-22-1 Equipment Drain

~11)-

V-22-2 12)

V-22-28 Floor Drain Sump 13)

V-22-29 14)

_V-31-2 Reactor Head Cooling 15 V-11-34 Isolation Condenser 16 V-11-36

17) NSO4A-Position A,B,D Limit Switch Main Steam (MSIV)

18) NSO48-Position A,B,0 19.NS03A-Position A,B,0

- 20 NS03B-Position A,B,D 21)

V-23-17 Solenoid Valve Nitrogen Purge 22)

V-23-18 23 V-23-19

- 24 V-23-22 25 V-23-20 26)

V-23-21 27)

V-28-17 Torus Ventilation 28)

V-28-18 z-29)

V-28-47

- 30)

LS-RD-87C (RD-08)

Level Switch Scram Discharge Volume 31)

LS-RD-88B (RD-08) 32)

LS-RD-91A (RD-08) 33)

LS-RD-920 (RD-08)

Plant ID#

Component-System

,s -

34)

V-14-30' Motor Operated Valve. ' Isolation Condenser 35)

V-14-31 136)

V-14-32 37)

'V-14-33 38)

V-14-34 39)

V-14-35 40 V-14-36.

41 V-14-37 42

_V-16-1 Clean Up (Isolation Valve) 43).

'V-16-2 44)

V-16-14 45)

V-17-1 Shutdown Cooling (Isolation Valve)

.46)

V-17-2 47)

V-17-3 48 V-17-19 49 V-17-54 D-050le

,