ML20009A650
| ML20009A650 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/30/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-06-10.A, TASK-6-10.A, TASK-RR LSO5-81-06-136, LSO5-81-6-136, NUDOCS 8107130433 | |
| Download: ML20009A650 (17) | |
Text
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June 30, 1981 Docket fio. 50-219 g
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N 101981w Mr. I. R. Finfrock, Jr.
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A, Vice President - Jersey Central
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Power & Light Company y
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Post Office Box 388 Forked River,llew Jersey 08731 ro
Dear Mr. Finfrock:
SUBJECT:
SEP TOPIC VI-10.A, TESTIf4G OF REACTOR TRIP SYSTEM AfiD EllGIllEERED SAFETY FEATURES (0YSTER CREEK)
A copy of our contractor's draft evaluation of Systematic Evaluation Program Topic VI-10.A 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.
In addition to correcting any errors that may exist in our evaluation you are requested to provide responses to the enclosed request for additional information within this same 30 day period.
This evaluation will be a basic input to the staff's safety evaluation report for this topic 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 fiRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, kb[dochofg8gjf, Dennis M. Crutchfield, Chief P
PDR Operating Reactors Branch #5 Division of Licensing nclosures:
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As stated 0,
cc w/ enclosures:
I See next page f's s : n sI ORB #5:D,
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OFFICIAL RECORD COPY "o ' + W
Mr.1. R. Finf rock, J r.
cc G. F. Trcwbridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 GPU Service Corporation t
ATTN: Mr. E. G. Wallace Comissioner Licensing Manager New Jersey Department of Energy 260 Cherry Hill Road 101 Comerce Street Parsippany, New Jersey 07054 Newark, New Jersey 07102 Natural Resources Defense Council Plant Superintendent 91715th Street, N. W.
Oyster Creek Nuclear Generating Washington, D. C.
20006 Station P. O. Box 388 forked River, New Jersey 08731 Steven P. Russo, Esquire 24S Washington Street Resident Inspector P. O. Box 1060 c/o U. 5. NRC Tors River, New Jersey 08753 P. O. Box 445 Forked River, New Jersey 08731 Joseph W. Ferraro, Jr., Esquire Deputy Attorney General Director, Criteria and Standards State of New Jersey Division Department of Law and Public Safety Office of Radiation Prograrn 1100 Raymond Boulevard (ANR-460)
Newark, New Jersey 07012 U. S. Environrental Protection Agency Ocean County Library Washington, D. C.
20460 crick Township Eranch 401 Chambers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region 11 Office Mayor ATTN:
EIS COORDINATOR Lacey Township 26 Federal Plaza P. O. Box 475 New York, New York 10007 Forked River, New Jersey 08731 Commi ssioner Department of Public Utilities State of New Jersey 101 Cornerce Street Newari, New Jersey 07102 r
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j SEP TECHNICAL EVALUATION TOPIC VI-10.A TESTING OF REACTOR TRIP SYSTEM AND I
ENGINEERED SAFETY FEATURES 1
OYSTER CREEK Docket No. 50-219 I
i June 1931 i
E. W. Roberts 6-16-81 t
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CONTENTS
1 1.0
.n a.......................................................
cm1T R,n.
no t i u.0 o
i 3.0 REACTOR TRIP SYSTEM................................................ 4 3.1 Description................................................... 4 z.2 tvaluation.................................................... 5 4.0 CONTAlhMENT SPRAY SYSTEM........................................... 8 4.1 Description................................................... 8 4.2 Evaluation....................................................
9 e.
5.0
SUMMARY
11
6.0 REFERENCES
12
..pt S ON W l.
Comparisons of Oyster Creek RPS instrument surveillance recuirements witn BWR Stancard Tecnnical Specification recuirements......................................................
6 2.
Containment spray anc associatec system surveillance recuirements.....................................................
10 1
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t e
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J SEP TECHNICAL EVALUATION I
TOPIC VI-10.A TESTING OF REACTOR TRIP SYSTEM AND ENGlhiERED 5AFETY FEAILRES OYSTER CREEK i
1.0 INTRODUCTION
Tne objective of this reviepy.is to cetermine if all Reactor Trip System (RTS) components, including pumos and valves, are incluced in component and system tests, if the scope and frequency of periccic testing is acequate, and if the test program meets Current licensing criteria. The review will also accress tnese same matters with respect to the Containment Spray System i
(CSS) as a typical example of all Engineered Safety Feature (ESF) systems.
1 i
2.0 CRITERIA General Design Criterion 21 (GDC 21), " Protection Systen Reliability anc Testability," states, in part, that:
Tne protection system shall :e cesignec to ;ermit periccic testing of its functioning anen tne reactor is in toeraticn, inclucing a capability to test cnannels ince:encently to cetermine failure anc iosses of recunaancy tnat may have occrrrec.
Regulatory Guide 1.22, "Periccic Testing of the Protection System AC!uation Functions," states, in Section D.l.a, that:
The periccic tests snould duplicate, as closely as practicacle, tne performance that is required of Ine actuation aevices in tne event of I
an accicent; I
l anc further, in Section D.4, states that:
When actuated equipment is not testec curing reactor operation, it should be snown that:
I
a.
There is no practicable system design that would permit operation of the actuated equipment without acversely affecting tne safety or operacility of the plant, D.
The probability that the protection system will fail to initiate the operation of the actuated equipment is, and can ce maintained, atieptably low without testing the actuated equipment during reactor operation, and c.
Tne actuated equipment can be routinely tested when the reactor is shut down.
IEEE Standard 338-1977, " Periodic Testing of Nuclear Power Generating Station Class 1E Power and Protection Systems," states, in part, in Section 3:
Overlap testing consists of channel, train, or load-group verification by performing indivicual tests on the various components and subsystems of the cnannel, train, or load grcup.
The individual component and subsystem tests shall check parts of acjacent sucsyst' ems, such that tne entire channel, train or icac group will be verifiec by testing of incivicual comoonents or suosystems."
ana in part in Secticn 6.3.4:
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Response time testing snail ce recairec only on safety systems or suosystems to verify tnat the response times are witnin the limits of 1
the overall response times given in the Safety Analysis Report."
Sufficient overlac snall be proviced to verify overall system response.
The response-time test shall include as much of each safety system, f rom sensor input to actuated equipment, as is practicable in a single test. Where the entire set of equipment from sensor to actuated ecuipment cannot be tested at once, verification of system response time shall be accomplished by measuring the response times of discrete 2
_e portions of the system and showing that the sum of t"a response times of all is within the limits of tne overali system re.;irement.
In acdition, the following criteria are applicaole to the ESF: General Design Criterion 40 (GDC 40), " Testing of Containment Heat Removal System,"
states tnat:
d The co'ntainment heat removal syste'm snall be designeo to permit appro-priate periodic pressure anpfunctional testing to assure:
a.
The structural and leaktignt integrity of its components.
b.
The operaoility and performance of the active components of the system.
c.
Tne operacility of the system as a wnole anc under concitions as close to the design as practical, the performance of the full operational secuence that orings the system into cperation, including operation of applicab.le portions of the protection systems, the transfer between normal and emergency power scurces, anc the operation of tne associatec cooling water system.'
Stancarc Review Plan, Section 7.3, A::pencix A, "Use of IEEE Stancarc 279 in Ine Review of tne ESFAS an Instrumentation and Controls cf Essential Auxiliary Supporting Systems," states, in Section ll.b, that:
Periodic testing should cuplicate, as closely as practical, the inte-grated performance required from the ESFAS, ESF systems, and their j
essential auxiliary supporting systems.
If such a " system level" test can be performed only during shutcown, the testing ccne curing power operation must be reviewed in detail. Check that " overlapping" tests do, in fact, overlap from one test segment to another. For example, closing a circuit breaker with the manual breaker control switch may not be adequate to test tne ability of the ESFAS to close tne breaker.
3 i
3.0 REACTOR TRIP SYSTEM (RTS) 3.1 Description.
The system is made up of two incepencent logic channels, eacn having subchannels of tripping devices.
Each subchannel has an input from at least one independent sensor, monitoring each of the crit-ical parameters.
The output of each pair of subchannels is ccmbined in a one-out-of-two logic:
that is, an input in either one or both of the independent subchan-nels will produce a logic cnannel trip.
Both of the other two subchannels are likewise combined in a one-out-of-two logic, independent of the first logic channel. Tne outputs of the two logic channels are combined in two-of-two arrangement so that they must be in agreement to initiate a scram.
An off-limit signal in one of the subchannels in one of the logic channels j
must be confirmed by any other off-limit signal in one of the subchannels of tne remaining logic channel to provide a scram.
During normal operation, all vital sensor and trip contacts are closec, anc all sensor relays are operated energized. The control rod pilot scram valve solenoids are energized, and instrument air pressure is apoliec to all scram valves. Wnen a trip coint is reacnec in any of the cnitorea carameters, a contact ccens, de-energizing a relay wnicn controls a ccntact in one of :ne two succnannels.
The cpening of a subcnannel con-tact ce-energizes a scram relay wnich opens a contact in the ocwer supply to the pilot scram valve solenoics suppliec cy its logic channel.
To this point, only one-half the events required to produce a reactor scram have occurred. Unless the pilot scram solenoids supplied oy the other logic channel are de-energized, instrument air pressure will continue to act on the scram valves and operation can continue. Once a single channel trip is initiatec, contacts in that scram relay circuit cpen and keep that circuit de-energized until the initiating parameter has returned within operating limits and the reset switcn is actuated manually.
It should be noted that each control rod has individual pilot scram solenoids for each channel and an incividual air-operated scram valve. A normally-closed switch is pro-vided in each logic channel pilot scram solenoid circuit. This allows each 4
1 rod to be manually scrammed (tested) by opening both logic -hannel switches ano de-energizing the pilot scram solenoids.
This type of est would pro-vide the required overlapping test of the RTS.
The parameters (sensors) wnich are recuired to initiate reactor scram are listeo in TaDie 1.
The condenser low-vacuum sensors'are connected to the RpS trip system and can initiate a scrpm. However, the only instruments included in th'is table are those recuirid to prevent exceeding the fuel cladding integrity limits during n.ormal~ operation or operational transients; as described in the plant FSAR and, listed,in Table 4.1.1 of the Oyster Creek Technical Specifications.
3.2 Evaluation.
The Oyster Creek RTS is designea to ailcw o'verlap-ping tests from actuating device througn the control rods. The cesign allows incividual channel tests frcm sensors tnougn pilot scram valves while the reactor is in Cperation and the overlapping red scram tests curing refueling.
Althcugn one or more rod scram valves may fail during reactcr operaticn, the cnannel tests will verify that no common mode fail-ure will occur and sufficient pilot valves, will operate to shut cown the reactor.
Tacie i shows tne present Oyster Creek RT3 instr; rent surveillance requirements, inclucing frecuency.
The table also shcas tne current licen-sing recuirements for General Electric boiling water reactors as listec in the Standard Tecnnical Specifications. The tests snCan only involve single channels testing (half-scram).
Table 4.1.2 of the Oyster Creek Tecnnical Specifications require a dual channel test (scram) witn the same frecuency as the instrument channel tests. However, this test is not performed.
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It should be noted that Technical S;ecification Tacle a.i.2 cces not require the main steam-line isolation valve closure parameter channel cali-bration, altnough the Oyster Creek Technical Specification recuirement in Section 2.3 requires that a 10% v61tr closure initiate scram.
a i s: -
ications for General Electric boiling The Standard Technir water reactors (page 3/4 3-1, r: 4,raph 4.3.1.2) require the logic system 5
1
/"'s function test and simulated automatic operation at least every 18 months, i
}
t Available information indicates that the overlapping system test is r.ot y.
perfurmed at Oyster Creer;. althcugh Table 4.1.2 of the plant Technical Specificaticns requires tre e.inimum Cual Channel (scram) test frequency to coincide with the instrumentation requirements.
TABLE 1.
COMPARISONS OF OYSTER CREEK RPS INSTP. MENT SURVEILLANCE REQUIRE-
.ENTS! WITH SWR STANDARD TECHNICAL DECIFICATION REQUIREVENTS
<gis /
t so Channel Channel Functiogal Channel c
Check Test" Calibratien Oyster Oyster Oyster Instrument Channel Creek STS Creek STS Creek STS Manua', scram NA NA Q*
Q NA NA High reactor pressured NA NA Q-M Q*
Q
< ~.
I nign crywell pressureC NA NA Q*
M Q*
O s.
Lo,. reactcr water level D
D 0*
M 0+
'ign.ater le'.cl in scr r NA NA P
C c:sc arge vclure
-irn 'adicticn in air stelm-n' W
W s
E
'.i$6 tar elC Avera:e oc.,er range r:ri ttr 5
W W
?
yes,A; Intermediate range ecnitor S
W W
U,
^,
f.i c.." ' t D
S'J "ain steam-line isolation valve NA NA Q*
'm R
clcsure.
Turpine trip scram NA NA Q*
M NA R
Gener'. tor load rejection scram NA Q*
0*
Reactor mode switch in snutcown
'M R
'! A g.s position'e
'E O
TABLE 1.
(continued)
,, 3, I
.% c. /
w FREQUENCY NOTATION Notation Frecuency Notation Frecuency 5
At least once per P.
At least cnce per 18 months shift D
At least once per NA
.yot appli~ cable ~
- 24. hc urs
+
W At least once per
{W
~ Prior te start up
, cays
/
M At least once per 50 Prior to shutdown month Q
At least once per Q*
Based en unsafe failure rate 3 months cata acd reliability analysis.
Not less than one-ronth or greater than tnree months.
A cualitative determination of acceptable operability by ocservat'en of a.
cnannel behavior during coeration. This cetermination shall incluce, shere l
'-s possible, comparisen cf the cnannel with other independent channels s,,,,'
measuring the same variacle.
b.
Injection of a simulated signal intc tne chanrel to verify its ;recer resocrs' incl;cing, wnere apslicau;e, alarm and/or t-ip i-itiatinc action.
c.
Acfustment of cnannel c;tput sucn tnat it -e:p:ncs, aitn accept-cie range anc accuracy, t o k n c.m calues of the parameter nhich the ch;nnel measures.
Calicraticn snali enccmpass the ent;re cnannel, inclucing ecuip-nert actuaticn, alarm, or trip.
d.
Not identified ';itn react.r trio systen ir Oyster Creed serveillance recuirements (Table 4.1.1).
Only active during given portion of operation cycle.
e.
f.
Nct requirec by Oyster Creek T.S. (Table 2.l..
II Or Novemcer i, 1973, JCP&L sutT.itted a sumnary ccmpariscn between tne Oyster Creek Nuclear Generating Station protection system and the requirements of Regulatory Guice 1.22.
This ccmoarison sncws that Oyster Cree < complies with the Regu'atory Guide except that bypass ccr,citicns are i
not individually and autcmatically indicatec tc tne reacter e;erator.
No f
system response-time measurements from sensor to actuated equipment are required by the Oyster Creek Technical Specifications.
Control rod response-time testing is required prior to start up after a major refueling outage with eight rods required to be tested after a reactor outage or scram.
Section 3.2, Reactivity Control, provides the rod scram insertion time requirements.
On page 3.2-5 of the Oyster Creek Technical Specifications, 200 milli-seconds is listed as the required maximum response time of sensor and associated RPS system to the start of control rod motion.
However, no response-time testing is required in the Technical Specifications.
4.0 CONTAINMENT SPRAY SYSTEM 4.1 Descriction.
The containment spray cooling system consists of two indepencent cooling loops, each capable of removing fission product cecay heat f rom the primary containment.
Each loop consists of a pair of spray heacers in the drywell vessel, fed by two containment spray pumps and two heat exchangers.
Two emergency service water pumps supply cooling water to the heat exchangers. Each of the containment cooling system loops craws suction frcm the base of the a0 sorption pool.
The system offers comolete recuncancy since eitner of the two system locos sucolies mere tnan adeouate cooling capacility. One ccntainment spray pump, one emergency service water cump, and two crywell soray valves in eacn icop are startec or opened automatically.
All otner valves are normally-open.
The backup pumps for eacn loop must be coeratec manually.
Containment spray is initiated by high drywell pressure and low-low reactor water level, each utilizing four independent sensors. Two channel logic with twc sensors of each variable in each channel are connectec in a one-out-of-two taken twice scneme which provides the necessary redunoancy for the automatic start circuitry.
4.2 Evaluation.
The containment spray system design allows testing up to a block valve at the drywell vessel for design flow conditions with the reactor in operation.
The circuit design also ensures that the system 8
will be returned to normal immediately if conditions so re gire. Valve operating circuits are such tnat it will not be possible tm have both independent containment cooling system loops in the test position at the same time. The service water pumps can be started and tested under design flow conditions at any time. Air can be used to assure lines and spray nozzles are not clogged when the reactor is shut down.
Table 2 shows the current testing requirements for the containment spray system and associated systens. The Oyster Creek Technical Specifica-tion surveillance recuirements (Taole 4.1.1.1) co not identify the instru-ment (channel) for each safety system. Therefore, it is possible that no (or incomplete) surveillance could be performed on sensors and/or logic cnannels for some safety systems. No direct comparison can be made with the Standarc Technical Specifications since no directly similar systems are used for SWR-5s.
However, the surveillance recuirements for the containment scray and associated systems agree with emergency coolir., systems in the Stancara iecnnical Specifications. No response-time testing is recuired by tne Oyster Creek Technical Specifications.
5.0
SUMMARY
The Oyster Creek Technical S;acification surveillance recuirements (Tacle 4.1.1) co not icentify tne instrument (cnannel) for eacn safety system. Therefore it is ;ossible that no (cr inccmolete) surve.illance could be performec on sensors anc/or logic cnannels for scme safety systems.
Reactor Trio System.
As discussec in Section 3, *.he Oyster Creek reactor trip system design and testing meets the current licensing criteria listed in Section 2 of this report, with the folicwing exceptions:
1.
The plant Technical Specifications do not cecuire a test of tne reactor mode switch and some instruments are not specifically identified for RTS system testing.
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TABLE 2.
CONTAINMENT SPRAY AND ASSOCIATED SYSTEM SURVEILLANCE REQUIREMENTS ^
Instrument Channels Channel Channel Functional Channel Functional Unit Check Test Calibration Low-low water level Daily Monthly Quarterly b
High drywell pressure (increasing) b High drywell pressure (decreasing)
Manual startb b
Timing Relays
~( auto transfer)
Soray System and Other Ecuioment Pump operacility Once/montn. Also, after major maintenance and prior to start up following a refueling outage.
Automatic Actuation Test Every three months.
Emergency Service Water System Pump cperacility Once/ month. Also, after major maintenance anc prior to start up following a refueling outage.
Automatic Actuation Tcst Every three months.
Auxiliary Electrical Power Automatic actuation of diesels Each refuelinc outace. Proposed and transfer emergency buses Technical Specification change.10 f rom offsite to onsite emergency power a.
Information from Tables 4.1.1, 4.1.2, and Section 4.4 of Oyster Creek Technical Specification Change e7, dated November 5, 1971.
b.
Not identified in T.S. Table 4.1.1 for surveillance.
10
e 2.
Although required in Table 4.1.2, no dual channt. (scram) test is performed by Oyster Creek.
(Operational problem prevent this test during reactor operation.)
T'e plant Technical Specifications do not recuire periodic system 3.
r response-time measurements (as assumed in FSAR) from sensor to actuated ecuipment.
/
Containment Soray System.
Th$OysterCreekcontainmentspraysystem is designed to permit complete, periodic, independent logic channel testing (GDC 22), and the design allows appropriate periodic pressure and functional testing (GDC 40).
The testing of the transfer between normal and emergency power sources, anc the operation of the emergency service water system provides reasonable assurance for availability and meets current licensing criteria.
However, the Oyster Creek Technical Specifications surveillance requirements require no periodic response-time measurements f rom sensor to actuatec equipment.
In addition, some instrumentation such as drywell pressure (increasing and decreasing), manual actuation, and timing relays iauto-transfer) are not includec in surveillance recuirements (Tatie 2).
6.0 REFERENCES
1.
General Design Criterion 21, " Protection System Reliability and Test-ability," of Appendix A, " General Design Criterie of Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
2.
Regulatory Guide 1.22, " Periodic Testing of the Protection Sys;om Actuation Functions."
3.
IEEE Standard 338-1977, " Periodic Testing of Nuclear Power Generating Station Class lE Power and Protection Systems."
4 General Design Criterion 40, " Testing of Containment Heat Removal Systems," of Appendix A, " General Design Criteria of Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
11
j 5.
Nuclear Regulatory Commissicn Standaro Review Plan, Section 7.3, Appendix A, "Use of IEEE Standard 279 in the Review of the ESFAS and Instrumentation and Controls of Essential Auxiliary Supporting Systems."
6.
Standarc Technical Specifications for General Electric Soiling Water Reactors (EWRs), NUREG-0123, Revision 2, Fall 1980.
7.
Oyster Creek Nuclear Power "lant No. 1, " Facility Description and Safety Analysis Report," Amendment 3, dated January 25, 1967.
S.
Tecnnical Specifications and Bases for Oyster Creek Nuclcor Power Plant Unit 1, Appendix A, to Provisional Operating License DPR-16, Amencments 1 through 15, dated February 2a, 1976.
9.
JCP&L letter EAGS-80-383 (Finfrock) to NRC, dated August 11, 1980.
- 10. JCP&L letter to NRC, " Supplement =6 to Amencment -68," dated November 1, 1973.
11.
Telecon, J. Knucel, Oyster Creek, D. Weoer, EG&G Idaho, Inc.,
March 31, 1981.
12
REQUEST FOR ADDITIONAL INFORMATION ON OYSTER CREEK TOPIC VI-10. A.
1.
With regard to Table 1 in the draft technical evaluation for Topic VI-10.A, please provide the technical justification for not testing on a frequency that is as conservative as the Standard Technical Specifications for each of the following:
(a) APRM.. Channel Checks
-r (b)
IRM - Channel Calibration (c) MSIV - Closure Channel Functionar Tests and Channel Calibration (d)
Turbine Trip Scram - Channel Functional Tests and Channel Calibration (e) Reactor Mode Switch in Shutdown Position - Channel Functional Tests 2.
Describe how the bypasses of channels and systems satisfy the require-ments of IEEE Std. 279-1971 Sections 4.13 and 4.20.
Identify and provide the technical justification for each bypass that does not satisfy the cited criteria.
3.
Identify and justify each safety channel that does not ct.nply with Section 4.10 of IEEE Std. 279-1971 by virtue of not testing for all of the minimum performance requirements of Section 3 (9).
4 With regard to Table 2 in the draft technical evaluation for Topic VI-10. A, please provide either item 1 or item 2 from question 1 above fcr each of the paraneters with a "b" superscript.
NOTE:
For responding to these questions, you may substitute the following sections of the 1968 trial use guide for the 1971 standard:
1968 1971 3(i) 3(9) 4.10 4.10 4.13 4.13 4.20 4.20
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