ML19326D718

Technical Evaluation of Electrical,Instrumentation & Control Design Aspects of Override of Containment Purge Valve Isolation & Other Engineered Safety Feature Signals for Fort Calhoun Nuclear Power Plant.

ML19326D718
Person / Time
Site:	Fort Calhoun
Issue date:	01/31/1980
From:	Broderick N, Debby Hackett, Radosevic J EG&G, INC.
To:
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ML19326D710	List: ML19326D718
References
SAN-L-809027, UCRL-15172, NUDOCS 8007030208
Download: ML19326D718 (15)

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ggg 1183-4145 Enorgy Faocouromonts Group TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND' CONTROL DESIGN ASPECTS _

OF THE  ;

OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SIGi.!ALS FOR THE l FORT CALHOUN NUCLEAR POWER PLANT (D OCK ET 50 -285) 1 JANUARY 1980 SAN - L 809027 '

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l Work Performed for Lawrence Livermere Laboratory under U.S. Department of Energy Contract No. DE-ACO8-76 NVO 1183.

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&EGnG Energy M Ocur:m:nto Grtup Sen Remin Cper:tl3n3 1183-4145 l

TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SIGNALS FOR THE  ;

FORT CALHOUN NUCLEAR POWER PLANT (DOCKET 50 -285) by D. B. Hacxt:.+

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• 1 Approved for Publication h '-

John R. Radosevic Department Manager This Document is UNCLASSIFIED

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'Nicnolas E! BrodericK Department Manager Derivative Classifier Work Performed for Lawrence Livermore Laboratorv under U.S. Department of Energy Contract No. DE-ACO8-76 '.VO 1183.

SAN RAMON OPi: RATIONS 2001 CLO CAQW CANYCN AC AO SAN AAMON, CALiFC ANia 9.aSe3

ABSTRACT This report documents the technic evaluation of the electrical, instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the Fort Calhoun nuclear power plant. The review criteria are based on IEEE Std-279-1971 requirements for the safety signals to all purge and ventilation isolation valves. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the U. S. Nuclear Regulatory Comission by Lawrence Livermore Laboratory. .

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FOREWORD This report is supplied as part of the Selected El ectrical ,

Instrumentation, and Control Systems Issues (SEICSI) Program being conducted .for the U. S. Nucldar Regulatory Commission, Office of Nuclear Reactor 'P.egulation, Division of Operating Reactors, by Lawrence Livermore Laboratory. Field Test Systens Division of the Electronics Engineering Department.

The U. S. Nuclear . Reg'ilatory Commission funded the work under an authorization entitled "Electrico' Instrumentation and Control System Support," B&R 20 19 04 031, FIN A-0231.

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

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

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TABLE OF CONTENTS Page

.1. INTRODUCTION .. .- . . . . . . . . . . . .. , , 1 i

' 2. EVALUATION OF FORT CALHOUN NUCLE *R POWER PLANT. . . . . 3 2.1 Review Criteria . . . . . . . . . . . . . 3

- 2.2 Containment Ventilation Isolation Circuits Design Description . . . . . . . . . . . . 4 I

2.3 ~ Containment Ventilation Isalation System Design Evaluation. . . . . . . . . . . . . . . 5 2.4 Other Engineered Safety Feature System Circuits . . 6

3. CONCLUSION $ . . . . . . . . . . . , , , , , , 7

- REFERENCES . . . . . . . . . . . . . . . . , , 9 t

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TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SIGNALS FOR THE FORT CALHOUN NUCLEAR POWER PLANT (Docket No. 50-285)

D. B. Hackett EG&G, Inc. , Energy Measurements Group, San Ramon Operations

1. INTRODUCTION j

Several instances have been reported where automatic closure of the containment ventilation / purge valves would not have occurred because the safGty actuation signals were either manually overridden or blocked during normal plant operations. These events resulted from procedural inadequacies, design deficiencies, and lack of proper management controls.

These e/ents also brought into question the mechanical operability of the cor.tairpent isolation valves themselves. These events were determined by the U. S. Nuclear Regulatory Commission (NRC) to be an Abnormal Occurrence

(#3-5) and were, accordingly, reported to the U. S. Congress.

As a follow-up on this Abnormal Occurrence, the NRC staff is reviewing the electrical override aspects and the mechanical operability as:ec s .of containment purging for all operating power reactors. On Never.ber 28, 1978, the NRC issued a letter entitled " Containment Purging Durir.g Normal Plant Operation"1 to all boiling water reactor (BWR) and m i

, , e crossurized water reactor (PWR) licensees. 'In a letter 2 dated DQcembe 28',

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'1973, a letter dated February 8, 1979, and a letter dated April 12, 1979, the Omaha Public Power District (OPPD), the licensee for the Fort Calhoun nuclear power plant, replied to the NRC generic letter. The licensee met with the NRC-in Washington, D. C. on August 23, 1979 and participated in a conference call on October 25, 1979 to further discuss the plant status.

This document addresses only the electrical, instrumentation, and control (Elac'l design aspects of the containment ventilation isolation (CVI) and ~ other engineered safety features (ESFs).

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2. EVALUATION OF FORT CALHOUN NUCLEAR POWER PLANT 2.1 REVIEW CRITERIA The primsry intent of this evaluation is to determine if the following NRC staff criteria are met for the safety signals to all purge and ventilation isolation valves:

(1) Criterion no. 1--Th e overriding

• of one type of safety actuation signal (e.g., radiation) must not cause the blocking of any other type of safety actuation signal (e.g., pressure) to the isolation valves.

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

key lock switches) are provided to facilitate adequate administrative controls.

(3) Criterion no. 3--The system-level annunciation of the overridden status is provided for every safety system impacted when any override is active.

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 initiate isolation'of the containment ventilation system. Specifically, contai nment high radiation, safety injection actuation, and containment high pressure should automatically initiate CVI. This is in conformance with Branch Technical Posi Section 6.2.4 of the Standard Review Plan.gion 6.4 or

" Ice following definition is given for clari ty of use in this evaluation:

Override: The signal is still present, and it is blocked in order to perform a function contrary to the signal.

(2) Criterion no. 5--The instrumentation and control systems provided to initiate CVI should be designed and qualified as safety-grade equipment.

(3) Criterion no. G--The overriding or resetting

• of the isolation actuation signal should not cause the automatic reopening of any isolation / purge valve.

2.2 CONTAINMENT VENTILATION ISOLATION CIRCUITS DESIGN DESCRIPTION Fort Calhoun nuclear power plant has two ESF trains which can cause isolation of the containment ventilation system. Train A controls the two inboard containment ventilation valves, and Train B controls the two oatboard iso! . ,lon valves. The initiating contacts for each train are described below:

(1) Automatic Contacts (a) Containment high-radiation (one out of five logic).

(b) Safety injection actuation.

(c) Containment spray actuation.

(d) Containment isolation activation.

(2) Manual Contacts Emergency operate switch (includes containment isolation actuation, safety injection actuation, and containment spray actuatien).

The CVI valves controlled by one train also have derived signals from the other train as backup isolation signals. There are no reset or override switches in the CVI system.

• Ine rollowing definition is given for clarity of use in this evaluation: "

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

When a monitored plant condition (or manual input) calls for isolation, electric power is lost to the slave relays (e.g. , "I"). Con-

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tacts of the slave relays open to remove electric power from the solenoid valves, causing the isolation valves to close.

The CVI valve-solenoid valves must remain energized in order for the isolation valve to remain open. The slave relay circuit contains a seal-in contact to Laintain electric power to it as long as a CVI signal is not present. With a CVI signal present, the valves will not remain open and cannot even be opened by their manual switch.

Clearing the initiating isolation signal will clear the CVI signal . However, the isolation valves will not and cannot automatically reopen. Their manual switch must be reinitiated to the "open" position in order to reopen them. ,

The containment high-radiation monitors have a reset switch, as defined in this report. The high-radiation signal must be manually reset to be cleared.

2.3 CONTAINMENT VENTILATION ISOLATION SYSTEM DESIGN EVALUATION In response to this issue, the operation of the containment i ventilation purge valves at the Fort Calhoun nuclear power plant is being limited to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year on an interim basis. Venting is )

being performed througu a two-inch vent line. Both the CVI and the con-tainment isolation signals will close the isolation valves on this two-inch vent line. The CVI actuation system contains no resets or overrides. We conclude that NRC staff criterion nos.1, 2, and 3 are satisfied.

The containment isolation activation signal is formed by an "0R" gate of several signals, including containment high pressure. Since CVI is l initiated by the four automatic contacts discussed in Section 2.2, we conclude that NRC staff criterion no. 4 is satisfied. -

r From the 'information provided by the licensee during the August 0

23, 1979 meeting and a follow-up telephone call on October 25,1979 with the NRC, the radiation-monitoring equipment at the Fort Calhoun nuclear power plant is not designed and qualified as safety-grade equipment. We ,

conclude that NRC staff criterion no. 5 is not satisfied.

Resetting of the actuation signal cannot cause the C'/I valves to automatically ' reopen. To reopen these valves, their switches must be reinitiated to the "open" position. We conclude that NRC staff criterion no. 6 is satisfied.

2.4 OTHER ENGINEERED SAFETY FEATURE SYSTEM CIRCUITS As part of this review and from information obtained from the licensee, it was determined that the design of the containment isolation actuation system, the safety injection actuation system, and the contain-ment spray actuation system are functionally similar to the CVI system.

None of these systems have an override feature. The safety injection actuation signal has a " reset" switch, as defined in this report. We conclude that the NRC staff criteria are satisfied.

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3. CONCLUSIONS * ,

The EI&C design aspects of ecntainment purge valve isolation and other ESF signals for_ Fort Calhoun nuclear power plant were evaluated using those design criteria stated in Section 2.1 of this report.

We conclude that with one excaption, the CVI system design meets the NRC staff criteria. The single exception is that, on the basis of the

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4- information provided, the radiation-monitoring equipment does not meet the NRC staff criterion for safety-grade equipment. We recommend that all of

, the instrumentation and control systems provided to initiate CVI should be designed and qualified as safety-grade equipment.

We also conclude that the other ESF circuit designs discussed meet the NRC staff criteria.

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REFERENCES- t

- 1. NRC/ DOR letter (A. Schwencer) to OPPD, " Containment Purging During Normal Plant Operation," dated November 28, 1978.

2. OPPD . letter (T. E. Short) to NRC (H. Denton, Attn: R. W. Reid),

" Docket 50-285," dated December 28, 1978.

3. OPPD letter (T. E. Short) to NRC (H. Denton, Attn: R. W. Reid),

" Docket 50-285," dated February 8,1979.

4. OPPD letter (T. E. Short) to NRC (H. Denton, Attn: R. W. Reid),

" Docket 50-285," dated April 12, 1979.

5. OPPD meeting with NRC in Warhington, D. C. , " Containment Purging During Normal Plant Operation," held August 23, 1979.

~ 6. U. S. Nuclear Regulatory Commission, Standard Review Plan,

" Containment Isolation System," NUREG 75/087, Rev.1, Section 6.?.4.

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DISTRIBUTION LIST LLL/Livermore EG&G/SR0 Lawrence Livermore Laboratory EG&G, Inc.

P. O. Box 808 P. O. Box 204

- Livermore, California 94550 San Ramon, California 94583 J. D. Attebery, L-255 C. E. Brown L. L. Cleland, L-155 E. K. Collins (4 copies)

W. Gieri, L-154 J. H. Cooper E. A. Lafranchi, L-151 0. B. Hackett (5 copies)

V. R. Latorre, L-156 D. H. Laudenbach H. C. Mcdonald, L-161

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B. G. Mayn R. P. Rumble, L-154 (5 copies) M. W. Nishimura G. St. Leger-Barter, L-154 B. M. Shindell N. L. Salmon, L-154 F. J. Tokarz, L-90 R. A. Victor, L-152 NRC B. Wilcox, L-53 (3 copies)

TIO File, L-53 (12 copies)1 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 LLL/ Nevada D. G. Eisenhut, MS-416.

B. K. Grimes, MS-238 Lawrence Livermore Laboratory G. C. Lainas, MS-416 North Las Vegas R. F. Scholl, MS-416 (2 copies) 316 Atlas Circle P. C. Shemanski, MS-416 (2 copies)

Las Vegas, nevada 89030 C. Tondi, MS-41G J. G. Ibarra , L-577 US00E/ TIC L..R. Peterson, L-577 W. E. Reeves, L-577 .

U. S. Department of Energy R. E. White, L-577 Technical Information Center (27 copies)

P. O. Box 62 Oak Ridge, Tennessee 37830 EG&G/ Idaho T. Abernathy (2 copies)

EG&3, Inc. US00E/NV00

P. O. Box 1625 Idaho Falls, Idaho 83401 U. S. Department of Energy Nevada Operations Office A. Udy P. O. Box 14100 Las Vegas, Nevada 89114 J. A. Koch R. R. Loux LMS/gys/Leslie #6 R. B. Purcell

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