ML20004A481

Forwards Override and Reset of Control Circuitry in Ventilation/Purge Isolation & Other Engineered Safety Feature Sys

ML20004A481
Person / Time
Site:	Duane Arnold
Issue date:	05/14/1981
From:	Carfagno S FRANKLIN INSTITUTE
To:	Butcher E NRC
Shared Package
ML112231555	List: ML112231554 ML20004A481
References
CON-NRC-03-79-118, CON-NRC-3-79-118 NUDOCS 8105180282
Download: ML20004A481 (1)

Text

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b j J. Franklin Research Center A Division of The Franklin Institute May 14, 1981 Gi United States Nuclear Regulatory Commission ep fy')4 Washington, D.C.

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Attention:

Mr. Edward J. Butcher, Jr.

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FRC Project C5257

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NRC TAC No. 10180

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Title:

FRC TER: Override and Reset of Control Circuitry in the Ventilation /

Purge Isolation and Other Engineered Safety Feature Systems - Duane Arnold Energy Center

Dear Mr. Butcher:

Enclosed is a Technical Evaluation Report which reviews the override and reset control circuitry in the ventilation / purge isolation system and other engineered safety feature systems of the Duane Arnold Energy Center nuclear power plant.

The report includes a description of the criteria established by NRC for the review, an evaluation of the specified systems for compliance with the stated criteria and conclusions drawn by FRC.

Very truly yours,

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'f S. P. Carfagno Project Manager SPC/JS/ih cc:

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R. Wichman

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J. T. Beard

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I TECHNICAL EVALUATION REPORT OVERRIDE AND RESET OF CONTROL CIRCUlTRY IN THE VENTILATION / PURGE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SYSTEMS IOWA ELECTRIC AND POWER COMPANY DUANE ARNOLD ENERGY CENTEIL_... _....

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'X)CKET NO. 50-331 NRC TAC NO. 10180 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03 79-118 FRCTASK 190 Preparedby Franklin Research Center Author: J. E. Kaucher The Parkway at Twentieth Street Philadelphia, PA #9103 FRC Group Leader: J. Stone

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i Prepared for I

Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: J. T. Beard May 14, 1981 This report was prepared as an account of work sponsored by an i

agency of the United States Government. Ncither the United States Government nor any agency thereof, or any of their err.ployees, makes any warranty, expressed or impiled, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

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. Franklin Research Center A Division of The Franklin institute i

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TECHNICAL EVALUATION REPORT OVERRIDE AND RESET OF CONTROL CIRCUITRY IN THE VENTILATION / PURGE ISOLATION AND OTHER ENGINEERED SAFETY FEATURE SYSTEMS IOWA ELECTRIC AND POWER COMPANY DUANE ARNOLD ENERGY CENTER NRC DOCKET NO. 50-331 NRC TAC NO. 10180 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03-79-118 FRCTASK 190 Preparedby Franklin Research Center Author: J. E. Kaucher The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader: J. scone Prepared for Nuclear Regulatory Commission Wedington, D.C. 20555 Lead NRC Engineer: J. T. Beard i

May 14, 1981 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or impiled, or assumes any legal liability or responsibility f@r aay third party's use, or the results of I

such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

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. Franklin Research Center A Division of The Franklin Institute The Bengmm Franklin Parkwey, PMa., Pa 19103(21S1448 1000

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TER-C5257-190 1

ABSTRACT This report documents the technical evaluation of the design of electri-i cal, instrumentation, and control systems provided in the Duane Arnold Energy Center to initiate automatic closure of valves to isolate the containment.

The evaluation was conductec in accordance with NRC criteria, based on IEEE Std 279-1971, for assuring that containment isolation and other engineered safety features will not be compromised by manual overriding and resetting of the safety actuation signals.

It was concluded that the electrical, instru-mentation, and control systems in Duane Arnold Erirgy Center partially con-r form with the NRC criteria.

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'a TER-C5257-190 FOREWORD This report is supplied as part of the Review and Evaluatior,of Licensing Actions for Operating Reactors being conducted by Franklin Research Center (FRC) for the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation, Division of Licensing.

The work was performed by FRC, Philadelphia, PA, under NRC contract No.

NRC-03-79-ll8.

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.i TER-CS257-190 CONTEN'IS Section Title Page I

1 INTRODUCTION 1

2 EVALUATION.

2 2.1 Review Criteria 2

2.2 Containment Ventilation System Design Description.

3 2.2.1 Caneralized System Design 3

2.2.2 Logic Circuits for Reset Seal-in and Trip 3

2.2.3 Individual Valve Control Circuits 4

2.3 Containment Ventilation System Design Evaluation 6

2.4 Other Engineered Safety Feature (ESF) System Circuits 7

2.4.1 Description of RER System Design 7

2.4.2 Evaluation of Other ESF Systems Design.

7 2.4.2.1 RHR System.

7 2.4.2.2 Other ESF S'ystems.

8 3

CONCLUSICNS 16 4

REFERENCES.

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LIST CF FIGURES

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Title Page

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PCIS Control Scheme.

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RER Control. Logic 13 4'

Containment Spray Valve Control Circuit.

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INTRODUCTTOM Several instances have been reported at nuclear power plants in which the containment ventilation / purge velves would not have autcmatically closed when required because the safety actuation signals were either overridden or blocked during normal plant operations due to procedural inadequacies, design deficiencies, and lack of proper management controls. These instances also brought into question the mechanical operability of the containment isolation valves themselves. The U.S. Nuclear Regulatory Commission (NRC) judged these instances to be Abnormal occurrences (#78-5) and were, accordingly, reported to the U.S. Congress.

As a follow-ap on these Abnormal occurrences, the NRC staff is reviewing the electrical override aspects and the mechanical operability aspects of contain-ment purging for all operating power reactors. On November 28, 1978, the NRC issued a letter entitled " Containment Purging During Normal Plant Operation" [1]* to all boiling water reactor (BWR) and pressurized water reactor (PWR) licensees. In a letter dated January 3, 1979 [2], the Iowa Electric Light and Power Company (IEL), the Licensee for Duane Arnold Energy Center (DAEC), replied to the NRC generic letter. On August 31, 1979 [3], IEL provided additional information pertaining to the NRC generic letter. On March 28, 1980 [4], the NRC requested that the Licensee provide additional information concerning electrical bypass and reset of engineered safety feature (ESF) signals for DAEC.

IEL made a partial response on May 5, 1980 15), which addres' sed only the containment purge va3ve electrical design, and submitted a supplement on June 17, 1980 [6], which analyzed their system in i

terms of the NRC criteria for ESF equipment and presented electrical schematics, system diagrams, and electrical data to verify compliance.

The present technical evaluation report, which reviews the IEL documentation deals only with the design of the DAEC electrical, instrume ntr.cion, and control (EIEC) components of the centainment ventilation isolation (CVI) and other engineered safety features.

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.Numbers in brackets refer to citations in the list of references, Section 4.

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EVALUATION 2.1 REVIEW CRITERIA i

The primary intent of this evaluation is to determine if the following NRC staf f criteria are met for the safety signals to all ESF equipments o Criterion 1.

In keeping with the requirements of General Design Criteria (G0C) 55 and 56, the overriding

• of one type of safety actuation signal (e.g., radiation) should not cause the blocking of any other type of safety actuation signal (e.g., pressure) for those valves that have no function besides containment isolation.

o criterion 2.

Sufficient physical features (e.g., key lock switches) are to be provided to f acilitate adequate administrative controls.

o criterion 3.

A system-level annunciation of the overridden status should be provided for every safety system impacted when any override is active.

(See NRC Regulatory Guide 1.47.)

Incidental to this review, the following additional NRC staff design criteria were used in the evaluations o criterion 4.

Diverse signals should be provided to 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 portion of safety injection actuation) should automatically initiate CVI.

o Criterion 5.

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

o Criterion 6.

The overriding or resetting + of the ESF actuation signal should not cause any valve or damper to change position.

In this review, Criterion 6 applies primarily to other related ESF systems, because implementation of this criterion for containment isolation has been reviewed by the Lessons Learned Task Force, based on the recommencations in NUREG-0578, Section 2.1.4.

Automatic valve repositioning

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

+ Reset: the signa 3 has come and gone, and the circuit is being cleared in order to return it to the normal condition.

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TER-CS257-190 signal for those valves that have no function besices containment isolation.

However, ten ESF actuated valves (MO-2000. MO-!*02, MO-2005, MO-1932, i

MO-2007, MO-1934, MO-2006, MD-1933, MO-2001, and MO-1903), which have functions in addition to containment isolation, are provided with control circuitry that allows tne bypassing of automatic ESF actuation (see Figures 4 and 3).

These valves are normally shut on RHR automatic initiation. They may f

te opened through the use of two or three manual switches, a local keylocked control switch (42/CS, Figure 5), and either S17 or S17 and o18 (keylocked) depending on the presence of an attomatic intiation signal and low reactor vessel shroud level (see Fagure 4).

Following these manual actions, the automatic initiation of RER will not cause these valves to close. Although not a litsral violation of Criterion 1, this situation has been identified for NRC staff evaluation with respect to acceptability.

Two of the three switches required to bypass the RHR valves identified above (S18 and local control switches) are keylock-type switches ar.d will support adequate adminstrative controls. In addition, system level annunuciation of this ecndition is provided. Consequently, Criteria 2 and 3 are satisfied for these ten valves. Criteria 2 and 3 do not apply to other RHR valves.

Criterion 4 does not apply to the RHR sv;;em.

The Licensee has indicated that the instrumentation and control systems provided to initiate the RER system are designed in accordance with IEEE Std 279 and use Seismic Category I equipment. Therefore, NRC staff Criterien 5 is i

satisfied in the RHR system at CAEC.

The overriding or resetting of any RHR actuation signal wit. not cause any valva or damper to change position. Therefore, it was concluded that NRC staff Criterion 6 has been satisfied in the RER system at DAEC.

2.4.2.2 Other ESF Svstems Equipment level bypasses are provided for several equipment items at DAEC which, if actuated following one safety actuation signal, will prevent a secend safety actuation signal from causing the equipment to take its O[FNnklin Research Center

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a.

Reactor Water Sample Valve (Inboard)

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Reactor Water Sample Valve (Outboard)

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N S pply Isolation Valve (Inboard)

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Loop A Containment Atmosphere Monitor System Isolation valve SV-8101A to SV-8110A f.

Loep B Containment Atmosphere Monitor SV-8101B to SV-B110B System Isolation Valve Although not a li'.eral violation of Criterion 1, this situation has been identified for NRC staff evaluation with respect to acceptability.

Operation of these bypasses is controlled by keylock switches and their activation is annunciated. Consequently, NRC staff Criteria 2 and 3 are I

satisfied for these eight valves.

Criterien 4 does not apply to ESF valves other than PCIS valves.

The Licensee has indicated that the instr 6 mentation and control systems are designed in accordance with IEEE Std 279 and use Seismic Category I equipment. Therefore, Criterion 5 is satisfied.

Sevacal equipment items not related to the PCIS or RHR systems will, as currently designed, move to their normal, pre-accident, position upon safuguard signal reset. DAEC has provided proposed rodifications to these control circuits (Attachment 3 to Reference 6) which will, when implemented during the 1981 refueling outage, prevent such repositioning upon reset. FRC has reviewed these system modifications and concurs that Lollowing their I

impleat7tation Criteria 6 will be satisfied. The valves which move to their normal, pre-accident, position upon safeguard signal reset are:

a.

Reactor Recirculatien Vump Discharge Bypass Valve (MO-4629) ts. Reactor Recirculation Pump Discharge Bypass Valve (MO-4630) c.

Auto Depressurization Valve icv-4400) d.

Auto Depressurization valve (SV-4402) e.

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Consensate Pump Discharge Isolation Valve (SV-2234) 1 1.

Steam Line Drain Isolation valve (SV-2410) l m.

Consensate Pump Discharge Isolation Valve (SV-2435) n.

Steam Line Drain Isolation Valve (SV-2411) j c.

Consensate Pump Discharge Isolation Valve (SV-2436) p.

Air to Steam Pressure Reducer Valve (SV-2033) q.

Air to Steam Pressure Reducer Valve (SV-2034) i r.

Air to Condenser Discharge to Suppression Pool on RCIC (SV-2037) s.

Air to Condenser Discharge to "uppression Pool on RCIC (SV-1966) 1 t.

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CONCLUSIONS The EIEC design aspects of ESF systems for Uuane Arnold were evaluated using staff design criteria.

It is concluded that the PCIS circuit design at AEC satisfies the NRC staff criteria for containment ventilation and purging operation with the exception of Criterion 4.

Satisfaction of Criterion 4 will require that a i

radiation detector which monitors containment (i.e., drywell or torus) activity be provided and used to automatically initiate primary contai. ment isolation.

Other ESF System Circuits 1.

RER System The RHR circuit design at DAEC satisfies the NRC staf f criteria with the exception of Criterion 1 for ten valves which have functions in addition to containment isolation. These ten valves (listed in Section 2.4.2.1) may be required to provide containment spray for pressure control of the containment atmosphere in an accident environment. Opening of the valves i

in question requires multiple switch (keylock-type) operation, and system level annunciation is provided. In view of the possible operational requirements, administrative controls, and annunciation, FRC concludes that no modification to the RER control circuit design is necessary.

2.

Other ESF Systems The eight valves listed in Section 2.4.2.2 satisfy the NRC staff criteria with the exception of Criterion 1.

However, because of operational requirements, these valves have functions in addition to containment isolation (i.e., post-accident reactor water sampling, nitrogen purge, and containment atmosphere sampling). Bypass of ESF actuation signals is via a keylock-type switch and is annunciated.

In view of the operational considerations, administrative controls, and annunciation provided, FRC concludes that no modification to these valve control circuits is necessary.

In the case of the 21 valves which will return to their normal, pre-accident, position upon safeguard signal reset, it is concluded that staff criteria will be satisfied upon ecmpletion of the circuit modifications identified in Attachments 2 and 3 to Reference 6.

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However, until the modifications are completed, appropriate l

administrative controls must to instituted to ensure'that all operators i

are aware of th.'s condition and operational procedures are established' which ensure that. these valves remain in their post-accident position-i upon system reset.

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REFERENCES 1.

NRC, Letter to all EWR and PWR licensees.

Subject:

Containment Purging During Formal Plant Operation 28-Nov-78 2.

L. Liu (IELPCO)

Letter to T. Ippolito (NRC)

Subject:

Containment Purging During Normal Plant Operations Iowa Electric Light tnd Power Company, 03-01-79 3.

L. Root (IELPCO)

Letter to T.

Ippolito (NRC)

Subject:

Containment Purging During Lormal Plant Operations Iow4 Ele:tric Light and Power Compan', 31-08-79 4.

T. Ippolito (NRC)

Letter to D. Arnold (IELPCO)

Subject:

Request for Additional Information - Containment Purge System Duane Arnold (TAC 10180)

NRC, 28-03-80 5.

L. Root (IELPCO)

Letter to T. Ippolito (NRC)

Subject:

Additional Infors.. tion Concerning the Electrical Design of the Concainment Purge Valves Iowa Electric Light and Power Company, 05-05 80 6.

L. Root (IELPCO)

Letter to H. Denton (NRC)

Subject:

Electrical Aspects of Engineered Safety Jeatures (ESF) and ESF Reset Controls Iowa Electric Light and Power Company, 17-06-80 4 -h MJ Franklin Research Center 4 >mma w rw r===ommme i