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EGG-EA-5725

JANUARY 1982

SYSTEMATIC EVALUATION PROGRAM, TOPIC VII-2, ESF

SYSTEM CONTROL LOGIC AND DESIGN, YANKEE NUCLEAR

POWER STATION

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' ' D. J. Morken,

U.S. Department of EnergyIdaho Operations Office * Idaho National Engineering Laboratory

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This is an Informal report intended for use as a preliminary or working document !

Prepared for the.

f U.S. Nuclear Regulatory ComissionUnder DOE Contract No. DE-AC07-761001570II FIN No. A6425-1 Q EGnG,a.n.p

8202000220 820201PDR ADOCK 05000029P PDr$ .

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INTERIM REPORT)

Accession No. |

Report No. EGG-EA-5725

Contract Program or Project Title:

Electrical, Instrumentation, and Control Systems Support for theSystematic Evaluation Program (II)

Subject of this Document:,

Systematic Evaluation Program, Topic VII-2, ESF System Control Logicand Design, Yankee Nuclear Power Station

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Type of Document:Informal Report

Author (s):

D. J. Marken

Date of Document:

January 1982

Responsible NRC Individual and NRC Office or Division:

Ray F. Scholl, Jr., Division of Licensing

This document was prepared primarily for preliminary or internal use. it has not receivedfull review and approval. Since there may be substantive changes, this docurr.ont shouldnot be considered final.

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EG&G Idaho, Inc.Idaho Falls, Idaho 83415

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Prepared for the*

U.S. Nuclear Regulatory CommissionWashington, D.C.

Under DOE Contract No. DE-AC07-76|D01570 ,

NRC FIN No. A6425-1

INTERIM REPORT

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SYSTEMATIC EVALUATION PROGRAM

TOPIC VII-2ESF SYSTEM CONTROL LOGIC AND DESIGN

YANKEE NUCLEAR POWER STATION'

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| Docket No. 50-29

January 1982,

D. J. Morken

EG&G Idaho, Inc.

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ABSTRACT

This SEP technical evaluation for the Yankee Nuclear Power Station-

reviews the tpe of isolation devices used in the Engineered SafetyFeatures (ESF) systems, the isolation between ESF channels and the -

isol: tion of ESF systems from control and non safety s.vstems.

FOREWDRD ;

This report is supplied as part of the "51ectrical, Instrumentation,,

and Control Systems Support for the Systematic Evaluation Program (II)" !being conducted for the U.S. Nuclear Regulatory Commission, Office of !

Nuclear Reactor Regulation, Division of Licensing by EG&G Idaho, Inc.,Reliability & Statistics Branch.

The U.S. Nuclear Regulatory Commission funded the work under theauthorization B&R 20-10-02-05, FIN A6425-1.

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CONTENTS

1.0 INTRODUCTION .................................................... 1

2.0 CRITERIA ........................................................ 1

3.0 DISCUSS ION AND EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

3.1 General ................................................... 2

3.1.1 Emergency Core Cooling System ..................... 23.1.2 Cont ainment I sola tion System . . . . . . . . . . . . . . . . . . . . . . 43.1.3 Main Steam Isolation System ....................... 4

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4.0 SUMMARY ......................................................... 5.

5.0 REFERENCES ...................................................... 5

APPENDIX A--NRC SAFETY TOPICS RELATED TO THIS REPORT . . . . . . . . . . . . . . . . . 7

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SYSTEMATIC EVALUATION PROGRAM

TOPIC VII-2ESF SYSTEM CONTROL LOGIC AND DESIGN

YANKEE NUCLEAR POWER STATION

1.0 INTRODUCTION

The objective of this review is to determine if non-safety systemswhich are electrically connected to the Engineered Safety Features (ESF)are properly isolated from the ESF and if the isolation devices or tech-niques used meet current licensing criteria. The cualification of safety--

related equipment is not within the scope of this review..

Non-safety systems generally receive control signals from ESF sensorcurrent loops. The non-safety circuits are required to have isolationdevices to ensure electrical independence of the ESF channels. Operatingexperience has shown that some of the earlier isolation devices or arrange-ments at operating plants may not meet current licensing criteria.

2.0 CRITERIA

General Design Criterion 22 (GDC 22), entitled, " Protective SystemIndependence," requires that:

The protection system shall be designed to assure that theeffects of natural phenomena and of normal operating, main-tenance, testing, and postulated accident conditions onredundant channels do not result in loss of the protectionfunction, or that they shall be demonstrated to be accept-able on some other defined bases. Design techniques, suchas functional diversity or diversity in component design andprinciples of operation, shall be used to the tent practi-cal to prevent loss of the protection function

General Design Criterion 24 (GDC 24), entitled, " Separation of Protec-tion and Control Systems," requires that:

The protection system shall bE separated from Control sys-tems to the extent that failure of any single control systemcomponent or channel, or failure or removal from service ofany single protection system component or channel which isconnon to the control and protection systems, leaves intact.

a systen that satisfies all reliability, redundancy, and '

independence requirements of the protection system. Inter-~ connection of the protection and ccntrol systems shall be

limitedsgastoassurethatsafetyisnotsignificantlyimpaired

IEEE-Standard 279-1971, entitled, " Criteria for Protection Systems forNuclear Power Generating Stations," Section 4.7.2, states:

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The transmission of signals from protection system equipmentfor control system use shall be through isolation deviceswhich shall be classified as part of the protection systemand shall meet all the requirements of this document. Nocredible failure at the output of an isolation device shallprevent the associated protection system channel from meetingthe minimum performance requirements specified in the designb ases .

Examples of credible failures include short circuits, opencircuits, grounds, and the application of the maximtsu cred-ible AC or DC potential. A f ailure in an isolation deviceis evaluated in the same manneS as a failure f ther equip- .

ment in the protection system

3.0 DISCUSSION AND EVALUATION~

3.1 General .

The Standard Review Plan, Section 7.1-III defines Engineered SafetyFeatures (ESF) systems as those functions which are required to operate tomitigate the cgnsequences of a pcstulated accident. Yankee Rowe Technical

Core Cooling System (ECCS), (2)gineered safety features as (1) EmergencySpecifications identify the en

Containment Isolation System (CIS), and(3) Main Steam Isolation (MSI) System.

3.1.1 Emergency Core Cooling System.5 The function of the ECCS is,in the event of a loss-of-coolant accident, to inject borated water intothe reactor in sufficient avantity to limit fuel clad metal-to-water reac-tion to a negligible amount.

The ECCS is comprised of high pressure injection (HPCI) pumps, lowpressure injection (LPCI) pinps, a pressurized borated water accumulatorwith a nitrogen pressurization system and the associated valves and piping.

The ECCS is actuated by either of two redundant channels. Channel Aaction is initiated by low main coolant pressure monitored by the reactorprotective system (RPS) or by high containment pressure. Channel B actua-tion is initiated by low pressurizer pressure or by high containment pres-sure. Either channel can be initiated by manual switches on the maincontrol board.

The low main coolant pressure safety injection initiation signal isobtained from channel 1 of the three RPS main coolant pressure protective

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ch annels . Sensor MC-P/S-100 provides the signal for channel l current loop.The current loop is comprised of a signal conditioning device MC-PT-100, a <pressure indicator MC-PI-100, two bistables MC-B/S-100 and .MC-B/S-101, a -

isolator / repeater device and an isolation buffer with an output signal tothe safety panel display system (SPDS). Bistable MC-8/S-100 provides relaycontact outputs for high containment pressure reactor trip logic and closureof the non-return valves. Bistable MC-B/S-101 provides relay contact out-puts to the 104 main coolant pressure reactor trip logic and the low pres-sure safety injection system (SIS) train A logic. Pressure switch PS-239,

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monitoring vapor containment pressure, provides the second half of train A1SIS initiation. Either subchannel will initiate the relay logic circuits

of train A safety injection actuation systems (SIAS-A).

Initiation of channel B logic is from pressure switch PS-238 monitor-ing vapor containment pressure and from a pressure switch monitoring lowmain coolant pressure. Either switch will initiate the train B safetyinjection actuation system (SIAS-B).

Actuation logic in each channel is comprised of multi-pole Westing-house type W.L. relays and auxiliary relays. Contacts of these relays arecombined so that operation of either SIAS relay initiates pump and valvefunctions for HPCI and LPCI..

Both channels have auto-manual switches on the main control board (MCB)~

for manual bypassing SIS during normal plant depressurization. Individualcontrol switches on the SIS panel and the MCB panel permit manual actuationof individual pumps and valves.

Information of pump and valve status is from auxiliary contacts on pumpcontrols, actuation relays and valve position switches. Separate pressureswitches and pressure transmitters provide status indication and pressurerecordings in the control room.

Accunulator injection actuation is from the SIAS relays. Contactsfrom these relays, upon closing, initiate time delay relays TDC-SI-l andTDC-SI-2 in channel A and TDC-SI-3 and TDC-SI-4 in channel B. Closure ofboth TDC contacts in either channel energizes solenoid valves (NS-S0V-46 orNE-S0V-47) which vent the valve actuators, allowing the three pressuriza-tion valves to open. This initiates accumulator pressurization and safetyinjection flow through the normally open accumulator outlet valve MOV-SI-1.Level switches LS-SI-1, LS-SI-2, LS-SI-3 and LS-SI-4 monitor the accumula-tor liquid level. The level switches close on sensing low liquid level,energizing relays LSX-SI-1, LSX-SI-2, LSX-SI-3 and LSX-SI-4. Contacts fromthese relays energize solenoid valves 50V-SI-45, 50V-SI-56 and 50V-SI-57 toblock accumulator pressurization, vent the accumulator and close MOV-SI-1.

Control switches permit manual testing of MOV-SI-l and the solenoidvalves. Relay contacts provide status indication of the actuating relaysed position switches on the air operated valves indicate valve status.

Power for the logic trains is from the 125V DC panels. Channel A isfed from the 125V DC switchboard No.1 via MCB panel 6R and channel B fromthe 125V DC switchboard No. 2 via MCB panel IF. 'The three 480V AC emergencybuses each feed an HPCI and an LPCI pump. Emergency panel MCC No.1 feeds

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the four loop fill valves and emergency panel MCC No. 2 feeds the fou,r loopblock valves. Accumulator actuator solenoid valves S0V-SI-46 and 45 receiveo

power from battery No. I and 50V-SI-47 from battery No. 3. Accumulatorsafety actuation solenoid 50V-SI-56 is powered by battery No. I andSOV-SI-57 from ba ttery No. 3. Isolation of the 125V DC ECCS functions fromother functions on the same bus is by fuses and switch disconnect. Isola .tion of pumps and valves on the 480V AC buses is by circuit breaker.

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Ev al ua tion. The ECCS uses separate and redundant channels. Relay andswitch logic provides adequate isolation between the ESF systems, i;he RPS,and control and non-safety functions. Isolation of the ECCS main coolantflow actuation signal from the RPS is by relay contacts which is satisfac-tory. The use of separate power buses, breakers and circuit fuses providesadequate isolation between channels and from control and non-safety func-tions.

3.1. 2 Containment Isolation.7 The containment isolation system(CIS) is comprised of redundant check valves in each of the incoming pipelines and pneumatically and electrically actuated trip valves in each ofthe outgoing pipe lines used in reactor operation.

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The trip valve actuation system consists of two redundant actuationtrains. Each train is comprised of a pressure switch monitoring vapor con-

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tainment pressure (PS-CI-231 for channel A and PS-CI-230 for channel B), ;

Electro Switch Type LOR 125V DC relays, auxiliary relays, test-bypass.

switches and solenoid valves to actuate the trip valves. Closure of either i

pressure switch upon high vapor containment pressure (25 psig) actuatesthe LOR relays for that channel. Individual relay contacts from the LOR jrelay energize solenoid valves which vent pressure off the containmentisolation trip valves causing them to close. Either train A or train Bwill initiate valve containment isolation closure. Both trains must be

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reset to re-open the valves. A manual control switch for each logic train #

is located on the main control board and permits manual operation of eithertrain. Separate three position switches (test, normal and bypass) permittesting or bypassing of individual operation devices in either train. Thetest switch in the test position will actuate operating devices, closingthe selected valve. It requires the switches in both trains to be in thenormal or bypass position and individual controls to be reset to permitopening a valve. Auxiliary contacts on the LOR relays and the test-normal-

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bypass switches actuate annunciators to supervise the status of the logictrains and solenoids.

Power to train A logic circuitry and associated solenoid valves isfrom Battery No. 3, while train B logic circuitry is powered from BatteryNo . 1.

Vapor containment pressure is also monitored by pressure transmitterPT-CI-227, which is independent of the CIS circuitry.

Evaluation. Use of separate relays and switches in. each logic trainprovides adequate isolation between channels. Relay and switch contactsprovide adequate isolation from RPS, control and non-safety functions.

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3.1. 3 Main Steam Isolation System. The main steam isolation systems 'is also an RPS function, initiating reactor trip upon a two-out-of-three *

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i signal from three pressure switches on any one of the four main steam lines.| The thre

VII-1. A.g channel, two train logic is covered in detail is SEP topic| The two-out-of-three logic energizes intermediate relays KIA3

and KIA4 in train A logic and KIB3 and KIB4 in train B logic. Contactsfrom these relays initiate the trip signal to the four non-return valves(NRV) for steam line isolation.

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The trip relays may also be energized from relay contacts from thecontainment isolation relays, initiated from the high containment pressurelogic, CIS relays.

Valve limit switches monitor the valve positions and provide an MtVtrouble alarm. Pressure switches monitor low accumulator pressure, lowhy sulic pressure and low hydraulic level for each valve.

Power to the main steam isolation system logic is from the 125V DC bat-tery system. Batteries 1, 2 and 3 feed logic channels 1, 2 and 3. Linefuses are used to isolate the channel logic and train logic from other func-

j tions on the same power bus. Information was not available to determine the/ power source (s) to the NRVs.,<

Evaluation. The main steam isolation systems is comprised ofindependent pressure switches and relay logic which provides adequate

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isolation between channels and trains and isolation from control andnon-safety sys tems.

4.0 SLM1ARY

Based on current licensing criteria and review guidelines, the ESFsystem electrical circuits comply with all current licensing criterialisted in Section 2 of this report.

5.0 REFERENCES

1. General Design Criterion 22, " Protection System Independence," ofAppendix A, " General Design Criteria for Nuclear Power Plants," 10 CFRPart 50, " Domestic Licensing of Production and Utilization Facilities."

2. General Design Criterion 24, " Separation of Protection and ControlSystems," of Appendix A, " General Design Criteria for Nuclear PowerPlants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."

3. IEEE Standard 279-1971, " Criteria for Protection System 3 for NuclearPower Generating Stations."

4. Yankee Nuclear Power Station Technical Specifications Appendix "A" toLicense No. DPR-3, May 30,1978.

5. Yankee Atomic Electric Co. Drawings 9699-FM-83A, Rev. 9; %99-FE-5A,~

Rev.16; 9699-ESK-6AA, Rev. 5; 9699-ESK-6AB, Rev. 6; 9699-ESK-6AD,Rev. 0; %99-ESK-6C, Rev. 3; %99-ESK-11 A, Rev. 6; 9699-FE-lJ, Rev. 8-

and Wes tinghouse Drawing 601-J-880, Rev. 6. '

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6. SEP Topic VII-1.A. " Isolation of Reactor Protection System fromNon-Safety Systems," dated August 27, 1981.

7. Yankee Nuclear Power Station. License Application, FSAR Vol . l .January 3,1974. Yankee Atomic Electric Company Drawing 9659-FE-4F, |Rev.11; and Sketch 8105110218-02, Fig. 3.

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8. YAEC Elementary diagrams, NRV controis-sheet 1 of 4, YR-E-50-037 andsheet 2 of 4. YR-E-50-038, dated 3-13-81.

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APPENDIX A

PRC SAFETY TOPICS RELATED TO THIS REPORT

1. III-I " Classification of Structures, Components, and Systems"

2. VI-7. A3 "ECCS Actuation System"

3. VI-10.A " Testing of Reactor Trip Systems and Engineered SafetyFeatures, Including Response Time Testing"

4. VII-1.A " Reactor Protection System Isolation".

5. VII-3 " Systems Required for Safe Shutdown".

6. VII-4 " Effects of Failures of Nonsafety-Related Systems onSelected ESFs"

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