ML20064K121

Technical Evaluation of Electrical,Instrumentation & Control Design Aspects of Override of Containment Purge Valve Isolation & Other Engineered Safety Features for Kewaunee Nuclear Power Plant

ML20064K121
Person / Time
Site:	Kewaunee
Issue date:	09/30/1980
From:	Debby Hackett EG&G, INC.
To:	NRC
Shared Package
ML111750737	List: ML20064K121
References
EGG-1183-4168, NUDOCS 8010070320
Download: ML20064K121 (23)

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SEPTEMBER 1980

.,** NERGYMEASUREMENTS GROUP TECHNICAL EVALUATION OF THE ELECTRICAL, INSYRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SIGNALS FOR THE KEWAUNEE NUCLEAR POWER PLANT (DOCKET No. 50-305) l

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o TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE E

OVERRIDE OF CONTAINMENT PURGE VALVE,150LATION k

AND OTHER ENGINEERED SAFETY FEATURE SIGNALS

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KEWAUNEE NUCLEAR POWER PLANT i

l coocxaT No.50-306) by D. B. Hackett/B. Kountanis Approved for Publication l

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" J. R. Radosevic Department Manager This document is UNCLASSIFIED 1[MfgMJ ss qienolas E. 5roderick Department Manager l

Work Performed for Lawrence Livermore National Laboratory under U.S. Department of Energy Contrect No. DE ACQS-76 NVO 1183.

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t DISCLAIMER This report was prepared.as, an account of work sponsored by the United States Government.

Neither the United States nor the United States any of their employees, makes any warranty, of Energy, nor Department express or implied, or assumes any legal liability or responsibility for the accuracy, completeness cr usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe on privately owned rights.

Reference herein to any specific comercial product, process, or service by trade name, mark, manufacturer, or other-wise, does not necessarily constitute or imply its endorsement, recommend-j ation, or favoring by the United States Government or any agency thereof.

The views and opinions of authors expressed herein do r.ot necessarily state or reflect those of the United States Government or any agency thereof._

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l ABSTRACT E

This report documents the technical evaluation of the electrical, E

instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the f.

Kewaunee nuclear power plant. The review criteria are based on IEEE Std-279-1971 requirements for the safety signals to all purge and venti-lation isolation valves.

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FOREWORD This report is supplied as part of the Selected E'lectrical, Instrumentation, and Control Systems Issues (SEICSI) Program being con-e ducted for the U. S.

Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Operating' Reactors, by Lawrence Livermore Laboratory, Field Test Systems Division of the Electronics Engineering Department.

The U. S. Nuclear Regulatory Commission funded the work under an E

authorization entitled "Elec crical, Instrumentation and Control System Support," B&R 20 19 04 031, FIN A-0231.

The work was perfomed by EG&G, Inc., Energy Measurements Group, San Ramon Operations, for Lawrence Livermore Laboratory under U.

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Department of Energy contract number DE-AC08-76NV01183.

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l TABLE OF CONTENTS I

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1.

INTRODUCTION.

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EVALUATION OF KEWAUNEE NUCLEAR POWER. PLANT 3

2.1 Review Criteria 3

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2.2 Containment Ventilation Isolation Circuits Design Description 4

2.3 Containment Ventilation Isolation System Design J

Eval uation.

6 2.4 Other Engineered Safety Feature System Circuits 8

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CONCouSIONS.

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REFEaENCES.

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TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS 0F THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION AND

- E OTHER ENGINEERED SAFETY FEATURE SIGNALS FOR THE KEWAUNEE NUCLEAR POWER PLANT (Docket No. 50-305)

D. B. Hackett/B. Kountanis 4

EG&G, Inc., Energy Measurements Group, San Ramon Operations y

1.

INTRODUCTION Several instances have been reported where automatic closure of l

the containment ventilation / purge valves would not have occurred because the safety actuation signals were either manually overridden or blocked I

during normal plant operations.

These events resulted from procedural inadequacies, design deficiencies, and lack of proper management controls.

These events also brought into question the mechanical operability of the i

I containment isolation valves theorelves.

These events were determined by the U. S. Nuclear Regulatory Comission (NRC) to be an Abnormal Occurrence l

(#78-5) and were, accordingly, reported to the U. S. Congress.

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As a follow-up on this Ahnomal Occurrence, the NRC staff is reviewing the electrical override aspects and the mechanical operability j

aspects of contairment purging for all operating power r'e' actors.

On l

November 28, 1978, the NRC issued a letter entitled " Containment Purging i

During Nomal Plant Operation" [Ref.13 to all boiling water reactor (BWR) l and pressurized water reactor (PWR) licensees.

In a letter [Ref. 23 dated January 3,1979, the Wisconsin Public Service Corporation (WPSC), the licensee for the Kewaunee Nuclear Power Plant, replied to the NRC generic letter.

A plant visit was made on June 6,1979 by an NRC staff mener.

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accompanied by Lawrence Livennore Laboratory and EG8G, Inc. (San Ramon Operations) personnel.

The applicable drawings were reviewed with the planf. personnel, and the equipmcat was examined.

Additional ' nfonnation i

was requested by the NRC in a letter [Ref. 3] dated September 14, 1979.

WPSC replied in a' letter [Ref. 43 dated October 18, 1979 and described i

s design changes that they intef#d to make.

j This document addresses only the electrical,'instrumentstion,, and l

con l.rol (EISC) design aspects of the containment ventilation isolation (CVI)pnd other engineered safety features'(ESF's).

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EVALUATION OF KEWAUNEE NUCLEAR POWER PLANT

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2.1 REVIEW CRITERIA The primary intent of this evaluation is to detennine that the i

h following requirements are met for the saf6ty signals to all ESF equipment.

E (1)

Criterion no.1--In keeping with the requirements of l

GDC 55 and 56 [Ref. 53, the overriding

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safety actuation signal (e.g., radiation) should not E

cause the blocking of any other type of safety actu-ation signal (e.g., pressure) for those valves that have no function besides containment isolation.

(2)

Criterion no. 2--Sufficient physical features (e.g.,

l keylock switches) are to be provided to facilitate adequate administrative controls.

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Criterion no. 3--The system-level annunciation of the overriden status should be provided fcr every safety system impacted when any override is active (see R.G.

1.47).

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Incidental to this review, the following additional NRC staff design criteria were used in the evaluation:'

(1)

Criterion no. 4--Diverse signals should be provided to E

initiate isolation of the containment ventilation system.

Specifically, containment high radiation, safety injection actuation, and containment high pressure (where containment high pressure is not a E

portion of safety injection actuation) should auto-matica11y initiate CVI.

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• The following definition is given for clarity of use in this evaluation:

E order to perform a function contrary to the signal.

Override: The signal is still present, and it is blocked in E

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(2)

Criterion no. 5--The instrumentation and control systems provided to initiate the ESF should be,de-4 signed and qualified as safety-grade equipment..

(3)

Criterion no. 6--The overriding or resetting

• of the ESF actuation signal should not cause any valve or damper to change position.

Criterion 6 in this review applies primarily to related ESF systems because implementation of this criterion for containment isolation systems will be' reviewed by the Lessons Learned Task Force, based on the recommendations in NUREG 0578, Section 2.1.4 (Ref. 6].

Automatic valve

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repositioning upon reset may be accgtable when containment isclation is not involved; consideration will be given oc a case-by-case basis. Accept-a5111ty would be depend *ent upon system function, design intent, and suit-able operating procedures.

2.2 CONTAINMENT VENTILATION ISOLATION CIRCUITS DESIGN DESCRIPTION Kewaunee Nuclear Power Plant has two ESF trains which can cause isolation of the containment ventilation system.

The initating contacts for each train are combined. as parallel inputs to form an "0R" circuit.

These contacts are described below:

(1)

' Automatic Contacts i

(a)

Containment high radiation (two nomally-open contacts in parallel from the radiation moni-tors).

(b)

Safety injection actuation (a norm. ally-open contact).

"The following definition is given for clarity of use in this evaluation:

Reset: The signal has come and gone, and circuit.is being clear-ed in order to return it to the nomal condition.

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(2)

Manual Contacts (a)

Containment isolation pushbuttons (two~ nor.

mally-open contacts in parallel).

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(b)

Containment spray push'uttons (two normally-open contacts in series).

Each train includes separate automatic and manual tnput "0R" gates, a latching relay, and a slave relay with c' ntacts in the control o

circuits of the pilot solenoid valves that control the CVI valves.

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" reset" switch and a " reset" seal-in rel.ay work in conjunction with the latching relays to provide the reset function.

The " reset" switch and its seal-in relay are contacted downstream of all.the automatic initiating 8

contacts.

The manual initiating contacts are connected to a point down-stream of the " reset" switch.

The " reset" switch for each train is an unprotected, simple, spring-loaded, pushbutton switch.

When a monitored plant condition (or manual input) calls for isolation, electric power is provided to operate the latching relay (type MG-6 relay) which, in turn, energizes its slave relay (e.g., V10X).

Con-tacts of the slave relay open to remove electric power from the solenoid valves causing the isolation valves to close.

When the " reset" switen is operated, the " operate" coil of the l

latching relay is de-energized, the " reset" coil of the latching relay is energized, and the " reset" seal-in relay is energized.

With the latching relay in the reset condition, the slave reley.is de-energized making elec-tric powee available to the solenoid-valve circuits.

The seal-in relay 1

will stay energized by power obtained through the contacts of the initiat-5 ing condition (e.g., high radiation, safety injection).-

The circuit design does not include provisions to annunciate the

" reset" or overridden status. Valve r3:ition lights (i.e., full-open/ full-closed) and individual valve control switches are provided.

The switches are three-position type with spring return to automatic from the open position and maintained contact in the closed position. i

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i 2.3 CONTAINMENT VENTILATION ISOLATION SYSTEM DESIGN EVALUATION 4

l The CVI signal has a " reset" that is more properly termed a by-1 pass or override, as defined in this review. This override, which contains l

i a seal-in relay, constitutes a system-level override which prevents reacti-

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vation of CVI by other automatic input signals as long as an' isolation j

signal continues uninterrupted. While in this override condition, none of the automatic safety signals can cause the containment ventilation / purge valves to close. When the last isolation signal is interrupted or c'leared, l

! the seal-in relay will drop out and allow a subsequent isolation input signal to generate a CVI. actuation signal and reclose the valves.

The manual CVI inputs are always active and available to the reactor operator, however.

During this review, we determined that this override design does l

not satisfy NRC staff criterion no.1.

However, the licensee has recently comnitted to modify their system design.

The system modifications the licensee has committed to perform include moving the automatic safety inj ection (SI) signal over to the manual input "0R" gate.

This leaves the high radiation automatic input (s) as the only signal that can be overridden.

An automatic SI input will cause CVI, even with an existing high radiation override. We conclude that this modification will satisfy the NRC criterion no.

1.

In addition, rather than overriding a high radiation alarm, the licensee intends to l

attempt to clear the alarm first by slightly upscaling the alarm setpoint.

This would allow a subsequent high-radiation-initiated CVI in the event that the radiation level increases during the time that the containment is being purged. This higher alam setpoint should be reviewed.

i The existing CVI signal override (" reset") function uses simple, metal-ringed, spring-loaded pushbutton switches in each train which are located on the sloping part of the control panel.

We conclude that this does not meet NRC staff criterion no. 2 regarding physical features that facilitate adninistrative controls.

However, with the above described -

mcdfications to be perfomed by the licensee, an inadvertent system-level.

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b.1'ock of CVI will not likely occur.

In addition, two " reset" buttons (one for each train) must be pressed in order to open the inboard and. outboard valves.

Finally, the purge and vent isolation valves will 'not! re'open automatically following the " reset" or clearing of a CVI; a second inten-(

tional (manual) act of opening the valves must be initiated.

Another safety feature would be the addition of plexiglass covers over the switches.

We conclude that NRC staff criterion no. 2 will be' satisfied with the incorporation of these modifications.

The presence of an override ("r'eset") of the CVI signal is not annunciated which does not satisfy NRC staff criterion no. 3.

However, the licensee's proposed modifications eliminate the ' possibility of a system-level override of the CVI.

Furthermore, the modifications will include an alann that annunciates when a high-radiation-initiated CVI is overridden.

We conclude that NRC staff criterion no. 3 will be satisfied with the incorporation of these modifications.

The CVI signal is generated in each train by a safety injection signal as well as by either of two containment high-radiation signals. The safety inj ection signal is a result of several diverse input signal s, including containment high-pressure. Hence, the CVI system design includes diverse actuation signals, and we conclude that NRC staff critarion no. 4 is satisfied.

Based on information from the licensee [Ref. 7{, the con-tainment high-radiation signal is provided by safety-grade equipment.

Thus, we conclude that NRC staff criterion no. 5 is satisfied.

Resetting the safety injection signal cannot cause the CVI system to reset, nor will it cause the automatic reopening of the containment ventilation / purge valves.

Clearing the CVI isolation signal requires manual operation of the " reset" pushbutton switch in sach train. Reopening of the valves further requires the manual operation of the individual ventilation / purge valve switches.

NRC staff criterion no. 6 is satisfied; however, that evaluation ~ will be performed by the Lessons Learned Task l

Force as discussed in Section 2.1.

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2.,4 OTHER ENGINEERED SAFETi FEATURE SYSTEM CIRCUITS o

It was determined during this review that the contaihment spray and containment isolation (for fluid systems) have a " reset" override circuit similar to that of the CVI.

The containment isolation and containment spray systems each has

' a single automatic actuation input signal.

Hence, we conclude that NRC staff criterion no.'1 is satisfied.

The " reset" switches for both trains of the containm'ent spray and containment isolation are simple, metal-ringed, spring-loaded push-button switches on the slopirig part of the control panel.

The design and location of these switches are judged to minimize inadvertent " reset."

Another safety feature would be the addition of plexiglass covers over the i

switches. With that modification, we conclude that NRC staff criterion no.

2 is satisfied.

The presence of an override (" reset") of the containment spray or' containment isolation is not annunciated.-

We conclude that NRC staff criterion no. 3 is not satisfied.

We recommend the installation of the appropriate system-level annunciation of the overridden system for every -

safety system impacted.

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Based on information from the licensee, the ESF equipment (in-cluding radiation monitors) is safety grade equipment which meets the 4

intent of IEEE 279-1968. Thus, we conclude that NRC staff criterion no. 5 is satisfied.

Initiation of SI will cause the closing of the.CVI and CI valves.

l After 90 seconds, the circuit design allows the SI actuation signal to be overridden (" reset"). Thus, the SI actuation input signal to these systems can be - removed.

While none of the CVI or CI valves will automatically reopen, these systems could be reset and the valves manually reinitiated to an open position.

This design meets the NRC staff criteria used in this

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' ' ev'aluation. Any further evaluation of this situation will be conducted by the NRC staff outside of this evaluation.

l When the CVI or CI circuits are " reset", none of the valves will automatically reopen.

We conclude that NRC criterion no. 6 is satisfied for these systems.

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CONCLUSIONS 1

The EI&C design aspects of containment purge valve isoiation and other ESF signals for the Kewaunee Nuclear Power Plant were evaluated using those design criteria stated in Section 2.1 of this report.

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We conclude that, with the CVI system design modifications the licensee has coninitted to perform, and with the' addition of plexiglass covers over the reset switches, the CVI system design meets the NRC staff criteria.

The evaluation of the CVI system with regard to criterion no. 6 will be perfomed by the Lessons Learned Task Force. The NRC should review the new radiation monitor alam tetpoint discussed in Section 2.3.

We conclude that, with the addition of plexiglass covers over the reset switches, the other ESF circuit designs discussed (containment spray and containment isolation) satisfy the NRC criteria, with one exception.

The single exception is that there is no system-level annunciation of the overridden condition, as discussed in Section 2.4.

We recommend that safety-level annunciation be provided.

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1 REFERENCES f

lI 1.

NRC/ DOR letter (A. Schwencer) to WPSC (E. James), "Containme' t Purging n

l During Nomal Plant Operation," dated November 28, 1978.

2.

WPSC letter (E. James) to NRC (A. Schwencer) " Docket 50-305, Contain-ment Purging During Nomal Plant Operations," dated January 3,1979.

3.

NRC/ DOR letter ( A. Schwencer) to WPSC (E. Mathews), no title, dated i

September 14, 1979.

4.

WPSC letter (E. Mathews) to NRC ( A. Schwencer) " Docket 50-305, Operat-ing License DPR-43, Containment Purge and Vent System," dated October 18, 1979.

5.

10 CFR 50, Appendix A, General Design Criterion.

6.

U.S. NRC, "TMI-2 Lessons Learned Task Force Status Report and Short-Tem Recommendations", NUREG-0578, published July,1979.

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7.

WPSC letter (Charles A.

Schrock) to NRC (Bob Licciardo), no title, dated March 27, 1980.

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DISTRIBUTION LIST I,

f LLNL/Livemore EG&G/SRO Lawrence Livermore National Laboratory EG&G, Inc.

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P. O. Box 808 P. 0. Box 204 v

Livemore, California 94550 San Ramon, California 94583 M. H. Dittmore, L-97 (2 copies)

Author (2 copies)

C. E. Br'own (4 ccpies)

B. G. Meyn M. W. Nisnia:ra i

LLNL/ Nevada NRC Lawrence Livemore National Laboratory U. S. Nuclear Regulatory Comission P. O. Box 45 Washington, D.C.

20555 hj Mercury, Nevada 89023 W. E. Reeves, L-577 (2 copies)

J. T. Beard, MS-416 D. G. Eisenhut, PS-528 g

G.

Lainas, MS-416 P. C. Shemanski, MS-416 i

USDOE/NV00 USDOE/ TIC U. S. Department of Energy U. S*. Department of Energy Nevada Operations Office Technical Infomation Center P. O. Box 14100 P. O. Box 62

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1 Las Vegas, Nevada 89114 Oak Ridge, Tennessee 37830 i

J. A. Koch T. Abernathy (2 copies)

I R. R. Loux R. B. Purcell i

ENCLOSURE 6 I '. CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.1.7 The containment purge supply and exhaust isolation valves may be open for safety-related reasons [or shall be locked closed]. The containment vent line isolation valves may be open for safety-related reasons [or shall be locked closed].

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

(For plants with valves closed by technical specification)

With one containment purge supply and/or one exhaust isolation valve open, close the open valve (s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(Fer plants with valves that may be opened by technical specifications) 1.

With one containment purge supply and/or one exhaust isolation or vent valve inoperable, close the associated OPERABLE valve and either restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or lock the OPERABLE valve closed.

2.

Operation may then continue until performance of the next required valve test provided that the OPERABLE valve is verified to be locked closed at least once per 31 days.

3.

Otherwise, be in at least HOT STANDBY within the next six hours and j

in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.1.7.1 The

-inch containment purge supply and exliaust isolation valves and the -

-incT vent line isolation valves shall be detennined locked closed I

at least once per 31 days.

4.6.1.7.2 The valve seals of the purge supply and exhaust isolation valves and the vent line isolation valves shall be replaced at least one per _ years.

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,,%> CONTAINMENT SYSTEMS 3/4 4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolution times as shown in Table 3.6-1.

APPLICABILITY: MODES I, 2, 3 and 4.

ACTION:

With one or more of the isolation valves (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:

a.

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or b.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or

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Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange; or d.

Be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isola-tion time.

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SURVEILLANCE REQUIREMENTS (Continued) s 4.6.3.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

a.

Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b.

Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

4.6.3.4 The containment purge and vent isolation valves shall be demonstated OPERABLE at intervals not to exceed months. Valve OPERABILITY shall be determined by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type,B and C penetration, the combined leakage rate is less than or equal to 0.60La.

However, the leakage rate for the containment purge and vent isolation valves shall be compared to the previously measured leakage rate to detect excessive valve degradation.

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