ML19339D026

SEP Review of NRC Safety Topic VII-2 Associated W/ Electrical,Instrumentation & Control Portions of Engineered Safety Features Sys Control Logic & Design.

ML19339D026
Person / Time
Site:	Dresden
Issue date:	11/30/1980
From:	St Legerbarter LAWRENCE LIVERMORE NATIONAL LABORATORY
To:
Shared Package
ML17193A707	List: ML17193A708 ML19339D024 ML19339D026
References
TASK-07-02, TASK-7-2, TASK-RR UCID-18698, UCID-18698-V3, NUDOCS 8102170038
Download: ML19339D026 (21)

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SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC VII-2 ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION AND CONTROL I PORTIONS OF THE ESF SYSTEM CONTROL LOGIC AND DESIGN FOR THE DRESDEN STATION, UNIT II NUCLEAR POWER PLANT Gerald St. Leger-Barter November 1980

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This is an informal report intended prunardy for laternal or thnated external disenbution. ,

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llte opmions and conctumons stated are those of the author and may or may not be thoes y; y; w.

of the bboratory. - s .<, , ,

5, This work was supported by the United States Nuclear Regulatory Cc.mmission under "&fW a Memorandum of Understanding with the United States Department o.' Energy. g. Nf$

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ABSTRACT This report documents the technical evaluation and review of NRC Safety Topic VII-2, associated with the electrical, instrumentation, and control portions of the ESF system control logic and design for the Dresden Station -

Unit II nuclear power plant, using current licensing criteria.

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FOREWCRD This report is supplied as part of the Systematic Evaluation Program being conducted for the U.S. Nuclear Regulatory Commission by Lawrence Livermore National Laboratory. The work was performed under U.S. Department of Energy contract number DE-AC08-76NV0ll83.

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TABLE OF CONTENTS Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . 1 _

2. CURR ENT LICENSING CRITERI A . . . . . . . . . . . . . . . 3

3. REVIEW GUIDELINES . . . . ............... 5

4. SY STEM DESCRI PTION . . . . . . . . . . . . . . . . . . . 7 4.1 Core Spray Subsystem . . . . . . . . . . . . . . . 7 4.2 Low Pressure Coolant Injection Subsystem . . . . . 8 4.3 High Pressure Coolant Injection Subsystem. . . . . 8 4.4 Automatic Pressure Relief Subsystem. . . . . . . . 9

5. EVALUATION AND CONCLUSIONS . . . . . . . . . . . . . . . 11

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SUMMARY

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l APPENDIX A NRC SAFETY TOPICS RELATED TO THIS REPORT. . . . . A-1 l

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F SYSTENATIC EVALUATION PROGRAM REVIEW 0F NRC SAFETY TOPIC VII-2 ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTIONS OF THE ESF SYSTEM CONTROL LOGIC AND DESIGN FOR THE DRESDEN STATION UNIT II NUCLEAR POWER PLANT ,

GERALD ST. LEGER-BARTER

1. INTRODUCTION The Engineered Safety Features Actuation Systems (ESFAS) of both PWRs and BWRs may have oesign features that raise questions about the electrical inoepenoence of redundant channels and isolation between ESF channels or trains. ,

Non-safety systems generally receive control signals from the ESF sensor current loops. The non-safety circuits are required to have isolation devices to insure electrical independence from the ESF cnannels. The safety objective is to verify that operating reactors have ESF oesigns which provide effective ano qualified isolation between ESF channels, and between ESFs and non-safety systems.

This report reviews the plant's ESF EI&C design features to insure that the non-safety systems electrically connected to the ESFs are properly isolated from the ESFs. This report also reviews the plant's ESFs to insure that there is proper isolation between redunoant ESF channels or trains, and that the isolation devices or tecnniques meet the current licensing critaria cetailed in Section 2 of tnis report. The qualification of safety-related equipment is not witnin tne scope of this report and is discussed in NRC Safety Topic III-12 and NUREG-0458.

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2. CURRENT LICENSING CRITERIA GDC 22, entitled " Protection System Independence," states that-The protection system shall be designed to assure that the effects of natural phenomena and of normal operating, maintenance, testing -

and postulated accident conditions on redundant channels do not result in loss of the protection function, or that they shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component cesign and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

GDC 24, entitled " Separation of Protection and Control Systems,"

states that:

The protection system shall be separated from control systems to tne extent that failure of any single control system component or channel, or f ailure or removal from service of any single protection system component or enannel wnich is acannon to the control and protecticn system leave intact a system satisfying all reliability, recundancy, and incependence requirements of the -

protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

IEEE Std-279-1971, entitled " Criteria for Protection Systems for Nuclear Pcwer Generating Stations," states in Section 4.7.2 that:

The transmission of signals from protection system equipment for control system use shall be througn isolation devices which shall 3

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be classified as part of the protection system and shall meet all the requirements of this document. No credible failure at the output of an isolation oevice shall prevent the associated protection system channel from meeting the minimum performance requirements specified in the design bases.

Examples of credible failures include short circuits, open circuits, grounos, and the application of the maximum credible a-c -

or d-c potential. A failure in an isolation device is evaluated in the same manner as a failure of other equipment in the protection system.

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3. REVIEW GUIDELINES The following NRC guidelines were used for this review:

Verify that the signals used for ESF functions are isolated from redundant ESF trains or channels. Review the schematic diagrams to ,

assure tnat tne wiring satisfies the functional logic diagrams in the FSAR or its equivalent (GDC 22).

Verify that qualified electrical isolation devices are utilized when reaunaant ESF trains or channels snare safety signals. Identify and cescribe the type of isolation device employed (GDC 22).

Verify that the safety signals used for ESF functions are isolated from control or non-safety systems. Identify and describe the type of isolation oevice employed (GDC 24 IEEE Std-279-1971, Section 4.7.2).

Verify that the logic does not contain sneak paths that could cause false operation or prevent required action as the result of operation of plant control.

Identify the related NRC Safety Topics in an appendix to the report.

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4. SYSTEM DESCRIPTION Means are needed to provide continuity of core cooling during those i postulated accident conditions where it is assumed that mechanical failures l occur in the primary system and coolant is partially or completely lost from l the reactor vessel. Under these circumstances core cooling is accomplisned by -

means of the emergency core cooling system (ECCS). The ECCS consists of two j

inoependent core spray subsystems, the low pressure coolant injection (LPCI) l subsystem, the high pressure coolant injection (HPCI) subsystem and the I automatic pressure relief suusystem.

4.1 CORE SPRAY SUBSYSTEM ,

The core spray subsystem consists of two independent spray systems each with its own pump, valves, and associated piping and instrumentation. The water source is conson to both systems and can be from the suppression pool in tne torus or, by appropriate valving, from the contaminated cemineralized water storage tr.nk.

! Initiation of the core spray subsystem occurs on signals indicating reactor low-low water level and reactor low pressure or high crywell l pressure. Low-low water level and high drywell pressure are each detected by four inoependent level and pressure switches connected in a form of 1

l one-of-two-twice logic array. Water injection can start when the admission valve is opened and when the reactor vessel pressure drops below pump i

discharge pressure (350 psig). Rated flow is sprayed over the top of the core at 90 psig in the reactor vessel. Opening of the aamission valves is accomplished only after the reactor pressure decays to approximately the design disenarge pressure of the pump, at which time the permissive signal to l open the valves is initiated by two pressure switches connected in a one-out-of-two logic array, i

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4.2 LOW PRESSURE COOLANT INJECTION SUBSYSTEM (LPCI)

The LPCI subsystem consists of two main subdivisions: one the LPCI system, and the otner the containment cooling system. The major equipment of the entire subsystem consists of two heat exchangers, four containment cooling service water pumps, four main system pumps, two drywell spray headers, a suppression enamber spray header, and associated valving, piping and instrumentation. .

The system pumps are activated on either a signal of reactor low-low water level and reactor low pressure or a signal of high drywell pressure similar to that received by the core spray pumps. The initiation signal also trips the recirculation pumps and supplies a start signal to the diesel generator which will provide power for the pump prime mover if normal auxiliary power has failed. The valves in the-high pressure part of the system are activated on a preset reactor low pressure signal similar to that of the valving on the core spray subsystem, thereby establishing a flow path.

4.3 HIGH PRESSURE COOLANT INJECTION SUBSYSTEM (HPCI)

The HPCI subsystem consists of a single steam turbine driving a multi-stage high pressure pump and a gaar driven single stage booster pump, valves, nigh pressure piping, water sources, and instrumentation. The turbine is driven with steam from the reactor vessel. Exhaust steam from the turbine

( is discnarged to the suppression pool. Suction for the HPCI pump is taken from the suppression pool.

Initiation of operation of the system is on a signal of either reactor low water level or nigh drywell pressure. These level and pressure switches are in a one-of-two-twice logic array similar to the reactor protection system. The system will automatically maintain reactor water level between low level and hign level if required flow is within the capacity of the pump and the reactor is not depressurized below 165 psia.

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{1 4.4 AUTOMATIC PRESSURE RELIEF SUBSYSTEM (ADS)

The automatic pressure relief subsystem is provided for backup of the HPCI subsystem and performs the functions of vessel depressurization for all small breaks. When the automatic pressure relief subsystem is actuated, the critical flow of stesn tnrough the relief valves results in a maximum energy removal rate with a corresponding minimum mass loss. Since the the automatic pressure relief subsystem coes not provide coolant makeup to the reactor, its .

function is only in conjunction with the LPCI or core spray subsystems as a Dackup to the HPCI.

Automatic actuation requires coincident indication of reactor water low-low level and drywell high pressure which is maintained for a period of 2 minutes. It also requires LPCI pressure switches to be made up and/or core spray pressure switches to be made up (i.e., their pump outputs to be above preset levels). There are two actuation chains and each circuit requires the same parameters for actuation.

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5. EVAULATION AND CONCLUSIONS The primary sensors for initiation of tbc ECCS functions were determined to De pressure and level switenes. These switenes and their associated ESF systems are tabulated in Table 5.1. It is noted that some of the switch functions are shared but that the isolation between functions is by the use of -

separate contacts from the relay operated by the pressure / level switch. The reference orawings also note the two switen functions are incorporated in the pressure switches operated by a common actuator. The parameters that initiate tne primary functica are listed but not all parameters which control the valve sequencing are shown as the isolation carried througn for them is similar to that for initiation.

The systems were reviewed in accordance with the review guidelines for isolation anc any apparent sneak paths for false operations or inhibition of operation. The isolation of each system from other functions is accomplished by use of separate sensor cwitches and/or separate contacts.

Based on the review of plant crawings it is concluded that the ESF systems are adequately isolated from control or non-safety systems and each other in accordance with tJe requirements of section 2 of this report.

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TABLE 5.1 ECCS ACTUATION SYSTEM PRIMARY SEN!OR SWITCHES FUNCTION RELAY RELAY ECCS LOGIC Orywell High Pressure P.S.

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2202-5 _ 902-32 - Core Spray

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lo32A ~ 1530-108 902-32 ' LPCI 1430-103A 902-32 - HPCI 2202-5 1632A (B) 2NO,2COM+ 2330-142 ~

2202-6 _ 902-33 - Core Spray 1o328 ~ lo30-208

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1430-1038 2202-6 2NO , 2COM = HPCI 1o328 2202-5 902-32 Core Spray lo32C 1530-109 902-32 ~ LPCI 1430-104A 2202-5 (8) 2NO, 2COM - 902-32 - HPCI

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1632C ~ 2330-143

_2202-6 - 902-33 - Core Spray 16320 ~ 1530-209 1430-1048 2202-6 2NO , 2COM = HPCI l

1o320 2202-5  ; 902-32 ADS 1628A 287-101A 2202-5 7 902-32 - ADS 1o288 287-102A 2202-6 - 902-33 __ 902-33 - ADS

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lo29A 287-10888 ~ 287-1018 2202-6 - 902-33 7 902-33 -

ADS Accv8 487-10988 287-1028 12 l

FUNCTION R ELAY RELAY ECCS LOGIC Drywell 1 PSI High Pressure P.S.

I 2202-5 _ 902-32 - LPCI -

1501-62A ~ 1530-134 -

2202-5 902-33 1301-028

- 1530-234

- LPCI 2202-5 902-32 1m01-62C

' 1530-199

- LPCI 2202-5 - 902-33 -

LPCI 1501-620 ' 1630-299 i

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i TION RELAY RELAY E REACTOR LOW WATER LEVEL LIS 2 902-32 = ADS gn,! 7 7,8  ; 14do-100A ,

= Core Spray 2 902-32 5,6  ; = HPCI 63-7 1530-103 N pg 2202-6 902-33 7,3 = = ADS 2e3-72B 1430-105B

= Core Spray

= HPCI 2202-o. _ 902-33  ; LPCI 5,6 ~ 1530-203 264-72B 902-32 = ADS 2 3-7

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= Core Spray 902-32 HPCI 2 7h '

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= LPCI 2202-6 _ 902-33 7,8 ~ 1430-106B ADS 263-720

= Core Spra.y

HPCI 2202-6 5,6  ; 902-33 *

LPCI 2o2-72D 1530-204 14

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FUNCTION RELAY RELAY ECCS LOGIC Reactor Lcw Level Insice

_ Shroud LITS 2202-7 902-32 -

LPCI

- 1530-110 263-73A l

2202-7  ; 902-33 ~

LPCI 263-738 1530-211 Reactor Low Pressure P.S.

2202-5  ; 902-32 -- LPCI

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2o3-52A 1530-150 2202-5 902 --

~ Core Spray i

203-52)I'2NC,2COM ~

- 1430-107A

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Core Spray 1430-129A 2202-6 -- 902-33 ~~ LPCI l 263-528 ~~ 1530-250 902-33  : Core Spray

_2202-6 2NC, 2COM coa-acs

- 1430-107B 902-33  :: Core Spray 1430-1293 15

FUNCTION RELAY RELAY ECCS LOGIC CORE SPRAY PUMP DISCHARGE P.S.

2202-19A _ 902-32 ~ ADS

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1430-1466A 1430-125A l

2202-198 _ 902-33 - AOS

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I430-la o6B 1430-1258 2202-19A _ 902-32 _

' 3 1930-146eC ' 1430-128A 2202-198 _ 902-33 C g3 1430-14660 1430-1288 e

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FUNCTION RELAY R ELAY ECCS LOGIC LPCI PUMP DISCHARGE P.S.

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( 1554 A 902-32 or > C C AUS 1530-198 2202-19A 15548 /

3 2202-198 1554C 902-33  ;

or k 1530-298 ADS 2202-198 15540 j 2202-19A 1554E l

902-32

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1530-168 2202-19A 1554F

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T 2202-198 1554H or (r ; 902-33 ADS 1530-258 2202-19B 1554J J

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6. St# WARY i

Based on the review of Dresden Station Unit II Plant drawings it is

concluded tnat tne isolation of the ESF systems satisfies the current t

j licensing requirements in section 2 of this report.

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REFERENCE

1. Code of Federal Regulations, Title 10, Part 50 (10 CFR 50),1979, Appendix A, (General Design Criteria).

2. Commonwealth Edison Company, Dresden Station Unit II Final Safety Analysis Report.

3. Dresden II Mechanical Drawings: M-25, January 1978; M-26-1, August 1977; M-27, April 1977.

4. Dresden II Electrical Drawings: 12E2429, September 1976; 12E2430, Feoruary 1977; 12E2435, February 1977; 12E2436, September 1976; 12E2437, September 1976; 12E2438, September 1976; 12E2461, September 1976; 12E2462, September 1976; 12E2627, September 1976; 12E2528, December 1976; 12E2529, December 1976; 12E2530, December 1976.

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APPEi40lX A

1. Topic VI - 7.A.3 " Testability and Operability of the ECCS Actuation System".

2. Topic VI - 10.A " Testing of RTS ano ESF including Response Time Testing".

3. Topic VI - 10.B " Shared ESFs On-Site Emergency Power and Service Systems for Multiple Unit Facilities".

4. Topic VII - 1.A " Isolation of the RPS from Non-Safety Systems".

5. Topic VII - 4 " Effects of Failure in Non-Safety Related Systems on Selected ESF's".

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