ML20126H135
| ML20126H135 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/30/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-07-02, TASK-7-2, TASK-RR LSO5-81-03-071, LSO5-81-3-71, NUDOCS 8104060696 | |
| Download: ML20126H135 (15) | |
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MAR 3 01981 ON Docket No. 50-219
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Mr. I. R. Finfrock, Jr.
Vice President - Jersey Central D
P Power & Light Company Post Office Box 388 U
it Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
SEP TOPIC VII-2, ESF SYSTEM CONTROL LOGIC AND DESIGN (0YSTERCREEK)
A copy of our current evaluation of Systematic Evaluation Program Topic-VII-2, ESF System Control Logic and Design is enclosed. This assessment compares your facility, as described in Docket No. 50-219, with the criteria currently used by the regulatory staff for ifcensing new facilities. Please infom us if your as-built facility differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment t
for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely.
Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
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MAR 3 01981 Docket No. 50-219 LS05-81-03-071 l
I Mr. I. R. Finfrock, Jr.
I Vice President - Jersey Central Power & Light Company Post Office Box 388
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Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
SEP TOPIC VII-2, ESF SYSTEM CONTROL LOGIC AND DESIGN (0YSTERCREEK) i A copy of our current evaluation of Systematic Evaluation Program Topic VII-2, ESF System Control Logic and Design is enclosed. This assessment compares your facility, as described in Docket No. 50-219, with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, kil DennsM.Crutchfield,CMef Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page
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i SEP 'rECHNICAL EVALUATION TOPIC VII-2
'ESF SYSTEM CONTROL LOGIC AND DESIG:1 OYSTER CREEK Doc 4et ilu. 50-219 March 1931 9
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CONTENTS
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1.0 INTRODUCTION
1 2.0 CRITERIA..........................................................
1 l
3.0 DISCUSSION AND EVALUATION...............................,,,,,,,,,
2 I
3.1 Discussion........................
2 i
j 3.1.1 Ceneral...............................................
2 i
3.1.2 Low Pressure Core Spray System 3
3.1.3 Containment Spray System..............................
4
- l 3.1.4 Automatic Depressurization System.....................
6 1
3.1.5 Emergency (Isolation) Condenser System'................
8 3.1.6 Coatainment Isolation System..........................
9 4.0
SUMMARY
10
~.0 REFERENCES 11 APPENDIX A--NRC SAFETY TOl'ICS RELATED. TO !!!IS REPORT...................
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SEP TECilNICAL EVAltlATION POPIC VII-2 ESP SYSTEM CONTROL LOGIC AND DES [CN OYSTER CREEK
1.0 INTRODUCTION
The objective of this review is to determine if non-safety systems which are electrically connected to the Engineered Safety Features (ESF)-
are properly isolated f rom the ESF and if the isolation devices cr tech-niques used meet current licensing criteria. The qualification of safety-related equipment is not within the scope of this review.
Non-safety systems generally receive cantrol sigaais f rota ESF senaar current loops. The non-safety circuits are required to have iavlatica devices to ensure electrical independence of the ESF channels. Operatia experience has shown that some of the earlier isolation devices or arran e-g meats at operating plants may not meet current Licensi.tg crits 2.0 CRITERIA General Design Criterion 2d (GDC 22), ent : t l a... " Protective System Independence," requires that:
The protection system shall c'e designed to na8ure tnat the effects of natural paenomena and of normal operating, maintananeo, tasting, and postulated accident conditions on redundant chanacts do not result in loss of the protection function, or that they shall be dectanstrated to be acceptable on some other defined bases. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
1 l-
j General Design Criterion 24 (GDC 24), entitled, " Separation of Protection and Control Systems," requires that:
i The protection system shall be acparated from control systems to the I
extent that failure of any single contro: system component or channel, t
.j or failure or removal from service of any single protaction system I
component or channel which is common to the control and protection f,
systems, leaves intact a system that satisfiev all reliability, redun-l dancy, and independence requirements of the protectiaa system.
Inter-f connection of the protection and control.;ystems shall b: limited so as to assure that safety is not significantly impairel.
IEEE-Standard 279-1971, entitled, " Criteria for Pratectica System: for Nuclear Power Generating Stations," Section 4.7.2. atates:
The transmission of signals from protection system equipmeat for con-trol system use shall be through isolation devices which ahall be classified as part of the protection system and snall meet all the requirements of this document. No credible failure at the output of an isolation device shall prevent the associated protectisn system channel from meeting the minimum performance requirement; specified in the design bases.
Examples of credible failures include short circuits, open circuits, grounds, and the application of the maximum credible AC
. JC poten-tial. A f ailure ia an isolation device ia evaluated ia tne same man-failure of other equipment in the protectica system.
ner as a 3.0 DISCUSSION AND EVALUATION 3.1 Discussion 3.1.1 General. The Oyster Creek FD&SAR does not specifically differentiate between the Reactor Protection System and the Engineered Safety Features (ESF) System. The Standard Review Plan, Section 7.1-III, defines ESF systems as those functions which are required to function 1
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automatically to mitigate the consequences of a postulated design basis event.
Based on the above definition, the following safety systems are clas-sified for evaluation as ESF systems:
1.
Low Pressure Core Spray System 2.
Containment Spray System 3.
Automatic Depressurization System 4.
Emergency Condenser System 5.
Containment Isolation.
3.1.2 Low Pressure Core Spr-System.
Discussion: The core spray system evntrol logic consists of two inde-pendent logic channels with two subchannela arranged in a one-au:-of-tuo-taken-twice coafiguration. Activation of the core s p ra:r is from four bistable delta pressure sensors, RE02A, C and B, D, measurir.; lou-low reac-tor water level, or four pressure switches RV46A, C and B, D, monitoring high drywell pressure. Redundant relay logic from these sensors in each channel initiates startup of both diesel generators and two core spray t
pumps. Failure of either or both spray pumps to start witnin a preset time period as monitored by pressure switches RV29A, C and B, D, sill initiate start-up of the redundant core spray pump (s).
Low core spray pump discharge pressure is monitored by separate pressure switches RV40A, C and B, D, which will initiate start-up of the core spray booster pumps.
Valve position indicators and annunciators for the core spray system are provided by torque and position switches located on each of the valves.
Bypasses and test circuits are by relay contact inserted in or around the ESF control logic circuita.
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Use of relay logic in redundant channels provides electrical isolation between channels of the core spray system and from other ESF, control, and l
non-safety systems.
l Power for the core spray system logic is from the 125V de buses.
Channels A and B receive power from the 125V de Panel D while Channels C and D receive power from the 125V de Panel F.
Isolation bet.nen Channels A l
j and B and from other safety systems on Panel D is oy thermal circuit l
breaker.
Isolation between Channels C and D and from other satety func-
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tions on Panel F is by thermal-magnetic circuit breaker. Mstsr operated valves for each channels are powered from separate power buses and are isolated from otner functions on the bus by magnetic-trip circuit breakers with motor-starter thermal-overloads. Howev4r, the thermal overloads for these circuits are jumpered.
Core spray pumps and booster pumps for eaca system are powered from separate power buses and are isolated from other power functions on the same buses by AK 50 breakers.
Evaluation: The low pressure core spray system logic function i.s redundant and provides adequate isolation oetween channels anc free cantrol and non-safety systems. Power to the logic circuits is frca separate buses and isolated by thermal-magnetic breakars.
Power for the redundant core spray pumps is from separate power buses and isolated from other systems by AK 50 'reakers.
Power to motor-operated (MD) valves is isolated between o
channels by separate buses; however, the iecist en circuit breakers ar2 magnetic trip only, which does not comply witn t.m criteria or Nuclear Regulatory Guide 1.75, Rev. 1, Section C-1.
6 3.1.3 Containment Spray System.
Discussion: Control logic for the containment spray system consists of two channels with two subchannels per enannel configured in a one-out-of-two-taken-twice logic.
Initiation of containment spray is from two sets of bistable sensors. One set of four sensors, IP15A, C and 3, D, monitors L
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high drywell pressure while contacts from relays 110A,'B, C, and D, actu-ated by core spray sensors RE02A, B, C, and D, monitoring low-low reactor l
water level. Containment spray actuation requires coincident high drywell i
pressure and reactor low-low water level signals.
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The system consists of two loops with two containment spray pumps and i
5 one heat exchanger per loop plus essential valves. One containment spray f
pump in each loop and the associated loop valves are energized automatt-cally on demand from the containment spray relay logic system. The secon-dary pumps can only be started manually. Time-delay relays in each loop logic circuit initiates startup of the emergency service water pump for condenser coolant water flow.
Decay of high drywell pressure to a preset va l s (1 psig) will shut down the containment spray system.
Control logic for actuation of the containment spray system is pro-vided by combining the sensor-actuated relays into a relay matrix which actuates the dual-channel initiation signals.
Isolation is maintained between channels and from other ESF, control, and non-safety systems by independent relay contacts.
Bypass and test circuits are by relay contacts inserted in and around the containment spray logic circuits. Valve positions, motor starter lights, and annunciators are indicated from individual position switches on the valves and from motor contactors.
Power for the system logic is from the 123V de Panel F for System I and Panel D for System II.
Line fuses in both positive and negativa le;n of ths logic circuitry isolate the logi: from other circuits on Panels D and F.
Isolation of those circuits from other safety circuits on Panel D is by enermal circuit breaker and on Panel F by thermal-magaetic circuit breaker.
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l Power for Loop 1 motor-operated valves is supplied from Motor Control ~
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Center (MCC) 1A21B and Loop 2 from MCC 1321B.
Isolation frota other safety functions on the samt bus is by thermal-magnetic circuit breaker.
l Loop 1 contain6ent spray pumps are fed from substation power Bus lA2 i
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and Loop 2 from substation power Bus IB2. Emergency service water pumps 1
l for Loops 1 and 2 are fed power from emergency switchgear ases ID and IC, c
l respectively.
Service water pumps for each loop receive power from unit substation buses lAl and 1A4 Thus, adequate power isciation is maintained between loops.
Isolation of pumps from other safety and non~JaIety loads on the same bus is by AK 50 circuit breakers.
Evaluation: The containment spray logic functions are by relay actu-ation, redundant and provide adequate isolation between channels and from control and non-safety systems.
Power to the logic circuits are from separ-ate 125V de buses and isolated by thermal breakers in Panel D-1 and thermal-magnetic breakers in Panel F.
Redundant containment spray pumps and emer-gency service water pumps are powered from separate buses and isolated from otner systems by AK 50 breakers.
Power to the MO valves is isolated between caannels by separate buses.
However, with the exception of containment spray inlet bypass valves and the pressure suppressiva chambat spray valves, wnich are protec ted by thermal-magne tic breakers, all other valves are protected my isolation breakers with magnetic trip only, whici. does not comply with the criteria of Nuclear Regulatory Guide 1.75, Rev. 1, Sec-tion C-1, 3.1.4 Automatic Depressurication System.
Discussion: The Automatic Depressurization. System (ADS) consists of five electromatic relief valves operating from high reactor pressure actu-ated by individual preset pressure switches, by operator manual actuation, or by ESF logic.
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ESF initiation of the ADS requires coincident actuation of three reac-tor monitor systems, each utilicing four sensors in a one-out-of-two-taken-twice logic plus a nominal time delay. The ADS monitors are comprised of:
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Iow-low-low reactor water level switches RE19A, B, C, and D; high drywell pressure using contacts from the core spray logic relays 115A, 3, C, and D actuated from pressure switches RV46A, B, C, and D; and high core spray boos ter pump pressure using contac ts - f rom tha core spray logic relays ll4A, 3, C, and D actuated from pressure switches RV40A, B, C, and 3 located on the spray booster pumps discharu:; lines.
Redundant relay and timer logie derivd from thue s e.w,c s : pens ta:
five cicctromatic relief valses in e r.ime-sequenud
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solenoid-initiated, ai r-a:t uated.
The two enannels are : n.u p = den t ' ano '
isolated from each other. Uu ?: relay logic previJos isulaLiua betw;a channels, from the core spray.systee,. control system, and u.,n,atety systems.
Valve position :s indiento. for each valve f rom " uni-wi ten u" na each solenoid actuator. Testing of tae valve act uation (wner. che react 6r it down) is by maaual-control switch.
Logic tasting coa;ists of applying pressure to the sensors.
No bypass of tha.\\DS system exists. Channal isolation from the redundant channel, from RPS enannels, and from nonsafety circuits is by relay contact.
Logic for each valve actuation circuit is separately fused to prevent loss of actua tion of more than one valve should a single saort circuit occur in the logic circuitry ar vai<e actuat?r.
Power to the relay-logic circuita and the solenoid-valve actuators is frca the 125V de Panels D and F.
ADS Valves NR108A, C and E are nominally powered f rom two thermal breakers on Panel D while Valves NR1083 and D are powered from two thermal-magnetic breakers on Panel F.
Redaadancy of power is also provided by loss-of power transfer relays in esca of tae two main logic circuits.
Evaluation: Tne ADS logic functions are reduadant and provide ade-quate isolation oetween channels and from other safety and centrol systems.
Power to the logic channels and for solenoid actuated valves is from the 125V de Panels D and F.
Power isolation from other functiuus on the.m panels is by thermal breaker on Panel D and thermal-magnetic breaker on Panel F.
This system complies with all current licensing criterin listed in Section 2 of this report.
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3 3.1.3 Emergency (Isolation) C,ndenser System.
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Discussion: The Emergency Condenser System consists'of two condensers and appropriate valves and support systems to provide redundant coolant
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loops operating with natural rceirculatian upon demand.
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Initiation of control logic for tne Emergency Condenser System is from I
reactor high pressure or reactor low-low water level and a noainal 15 second delay. Four reactor high pressure swi t c'ac a, RE15A, B, C, an t.
D and relay i
contacts actuated by the RPS low-low reaetor water level switches RE02A, B, C, and D are arranged in a one-out-of-two-taken-twi:n logic.
Sensing of hign flow by redundant flos switenes on raen loop indicates tuop piping failure and automatically signals closure at tae redundant votlet and inlet block valves in the damaged loop.
Tne Emert;e..:: Goa% ns;r Sfste.n u. cala..rea wit.. ut
.;... E detailed electrical contral drawing.
a.i tne
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- .. s ;.u t e d information and oa detailea revieu o f atn.:r L.
loci:, i- !s
- 4.3 n.me ! :na t relay logic is also used in the eme rs.nc;. candens.r lor.ic abem;.;
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valve con t rc,1, bypas s e s, and manu 1 vverride circuitr..... L. t.. ;. :. 1.
valve position indication and aaauncia..
are from torque an; iimit seitenes on tue v t. l vu.
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supposition that the enannel redundancf and isolati;r. irum nvn-safety systems is adequate.
All valves are normally open encept for una 125V de valve in each condensate return line.
The 125V de valves are powered frx.: MCC DC-1 and tne ac valves are powered from MCC 1AEJ.
Loss of MCC DC-1 ous sculd pre-l vent opening recirculation velves in bata L' ops.
Evaluation:
Based on reviews of the Emergency condenser System docketed inf onnation, PSI Diagram, and evaluatica o.'
ather E55 logic cir-cuitry, aut without ene cenefit of in-depta review of electrical, eleman-tary, or scaematic diagrams, it is suggested the logic circuitry isol:::iaa bacween cnannels and from control and ava-safet;. systems is adequate.
"cuar to tne ac MO valves is fed from MCC 1A32 and ta che redundant d a valve:,
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from de Bus D-1.
Loss of Bas DC-1 would prevent opening the loops to emer-gency condenser flow.
Power isolation at the power pancis is by therm &1-magnetic breakers.
3.1.6 Containment Isolation System.'O'11 Discussion: The Containment Isolation Svstem consists at redundant isolation valves placed in aeries on all pipes and ducts wai:n penetrate the reactor primary evntainment.
In each line, cne.-alve is located inside the containment vessel aad the secoad valz; loca.eu actside.
Systema included in the Containmeat Isolation System evaluation are:
1.
Main steam line valves 2.
Isolation condenser 3.
Cleanup system 4.
Shutdown system 5.
Drywell vent purge and sumps.
Containment isolation may be initiated oy manual control of each valve or upon demand by RPS requirements.
Initiation of RPS contral logic is from reactor low-low water level or high drywall pressure.l
Cicsure of main steam and main steam-line drain valves may also a: initic;ed oy high radiation level, low steam-line pressure, or main steau '.ine braak moni-tored by temperature sensors in the pipe tunac1.
The four reactor lou-low water level switches, RE02A, 3, C, and D, and the four high drywell pres-sore switches, RE04A, B, C, and D, are caca arranged in a one-out-of-two-taken-twice logic to provide redundant trip channels.
Isolation between enannels and from other RPS, ESF, control, and non-safety systems is by relay contacts.
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Testing of individual valve actuation is by manual control while sys-tem testing is performed by pressurizing the RPS sensors.
Bypasses and test circuitry are by relay contacts inserted in and around the control l
logic circuits. Valve position indication and annunciation are by torque l
and limit switches located on the valves.
Power for the system logic is from the RPS motor-generator sets (see SEP Topic VII-1.2).
Individual relay logic circuits are separately fused to further isolate individual RPS and ESF logic circuits.
The main steam-line isolation valves are air-actuated from redundant ac and de solenoids.
115V ac f rom VACP-1 actuates the ac solenoid valves while the de solenoids are actuated from the 125V de Panel F.
Other iso-lation valves are located both inside and outside the primary containment.
Valves located inside the containment are powered from the vital Bus MCC-1AB2. Valves located outside the primary containment are powered from the 125V de Bus DC-1.
L Evaluation: The Isolation Contanment System logic functions are A
redundant and provide adequate isolation between channels and from control and non-safety systems. Power to the logic circuitry is from the RPS buses.
Power to the contaimment MO valves inside the containment is from MCC 1AB2 and for the outside MO valves from the 125V de Bus DC-1.
Each valve circuit is isolated from other power circuits on the same bus"by thermal-magnetic breakers except for four ac valves, V-16-1, V-16-61, V-17-19, and V-17-54, which use magnetic trip only breakers which do not comply with the criteria of Nuclear Regulator Guide 1.75, Rev. 1, Section C-1.
4.0
SUMMARY
Based on current licensing criteria and review guidelines, the ESF systems logic circuits comply with all current licensing criteria listed in Section 2 of this report.
Based on current licensing criteria and review guidelines, isolation of power circuits to 23 ESF valves does not meet the criteria of IEEE 10
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Standard 384, Section 3, as amended by Regulatory Guide 1.75, Rev. 1, Section C-1.
5.0 REFERENCES
I 1.
General Design Criterion 22, " Protection System Independence," of Appendix A, " General Design Criteria of Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities."
l 2.
General Design Criterion 24, " Separation of Protection and Control Systems," or Appendix A,
" General Design Criteria of Nuclear Power f
Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
3.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
4.
Oyster Creek Nuclear Power Plant Unit No. 1, " Facility Description and Safety Analysis Report," Final Amandment 3, Part 1, Vol. 1, January 25, 1967.
5.
NUS Corp Drawing 5060F032, Rev. 1, and General Electric Drawing 718E644, Sheet 2, Rev. 11.
6.
General Electric Drawing 237E901, Sheets 1 and 2, Rev. 7.
7.
MRP Assceiates Drawing 1083-55-09, Sheet 1, Rev. A.
3.
Letter (Finfrock) ta NRC, June 24, is75, "0ystar Creek Ds:ket 50-2!9 (697), Compliance with Provisional Operating Licensee, snandment 3, Licensing Condition."
9.
Letter (Finfrock) to NhC, February 2, 1977, "0yster Creek Docket 50-219 (1090), S;pplemeat No. 8 (Rev. 3) to Application for a Fall-Tern Oper-ating License."
10.
Burns and Roe Drawing 3028-12.
11.
Oyster Creek Nuclear Power Plant Primary Containment Report by the Ralph M. Parson Company, Amendment 15, September 7, 1967.
12.
General Electric Drawing 237E566, Sheets 1, 2, 4, 5, 6, Rev. 15.
1 11 l
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APPENDIX A i
NRC SAFETY TOPICS RELATED TO THIS REPORT r
i 1.
III-l
" Classification of Structures, Components, and Systems" i
i 1
2.
VI-7.A3 "ECCS Actuation System" l
3.
VI-10.A
" Testing of Reactor Trip Systems and Engineered Safety f
Featura. Including Raspans.: Time Testing" 4
VI!-l.A "Reacter Protection System Isolation" 5.
VII-J Systems R2 quired iar Sai.e Shutdown" c.
VII-4 "E f f.ec ts o f Failure:, sf :i.asafety-Related Syste.as on i
Selected ESFs" f
B f
f r
!i I
t i
l i
I 1
12 1
h l-I
- - -....... - - -. -.