ML20054C994

Informs of Modified Position on Length of Time Reactor Protection Sys or ESF Actuation Sys Channel May Be Bypassed. Proposed Tech Specs Encl

ML20054C994
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/31/1982
From:	Clark R Office of Nuclear Reactor Regulation
To:	Cavanaugh W ARKANSAS POWER & LIGHT CO.
References
TAC-7089, NUDOCS 8204220225
Download: ML20054C994 (2)

Text

hb MAR 311982 DISTRIBUTION:

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Senior Vice President, Energy ACRS-10 am Supply Department I&E 9

anumanmeT, Arkansas Power & l.ight Company Gray File P. O. Box 551 PMKreutzer-3 4

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Little Rock, Arkansas 72203 RAClark

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Dear Mr. Cavanaugh:

RMartin This is to inform you of a modified staff position on the length of time a reactor protection system (RPS) or an engineered safety features actuation system (ESFAS) channel may be bypassed, Your Technical Specifications (TS) currently limit the inoperable channel bypass time to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If it can be verified that three protection system chennels in two-out-of-three logic are suf.~ie 2nt to satisfy all protection system criteria, then one chamel of a four channel system may be bypassed for a lengthy period of time with no degradation to safety. Action Statement 2 of the Enclosure 1 TS page (San Onofre - Unit 2 type) may be proposed in place of thc appropriate present action requirements. delineates the criteria which must be satisfied to ensure that three channels are sufficient for plant protection.

However, we believe that it is prudent that an inoperable channel be repaired and returned to service as quickly as practicable. Action Statement 2 of Enclosure 1 would allow the flexibility of continued plant operation with an inoperable channel in bypass where it is not feasible to repair the chtnnel (for example, if the failure is inside the containment in a location inaccessible during plant operation). The model TS, however, do place responsibility on the licensee to repair the channel as quickly as possible and in any event during the first cold shutdown after the failure has occurred.

If you have any questions on this subject, please contact your assigned I

NRC project manager.

l Sincerely, l

Original signed by Robert A. Clark l

Robert A. Clark, Chief l

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A,rkansas Power & Light Company cc:

Mr.. John Marshall Manager, Licensing Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203

'Mr. James P. O'Hanlon U.S. Environmental Protection Agency Region VI Office General Manager ATTN:

Regional Radiation Arkansas Nuclear One Representative P. O. Box 608 1201 Elm Street Russellville, Arkansas 72801 Dallas, Texas 75270 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220 7910 Woodmont Avenue S. L. Smith, Operations Officer Bethesda, Maryland 20814 Arkansas Nuclear Planning &

Response Program Nicholas S. Reynolds, Esq.

P. O. Box 1749 c/o DeBevoise & Liberman Russellville, Arkansas 72801 1200 Seventeenth Street, N.W.

Washington, D. C.

20036 Arkansas Polytechnic College Russellville, Arkansas 72801 Mr. Charles B. Brinkman Manager - Washington Nuclear Operations C-E Power Systems 4853 Cordell Avenue, Suite A-1 Bethesda, Mhryland 20014 Regional Administrator Nuclear Regulatory Commission, Region IV Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. W. Johnson i

U.S. NRC P. O. Box 2090 Russellville, Arkansas 72801 i

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a ENCLOSURE 1 TABLE 3.3-1 (Continued)

TABLE NOTATION With the protective system trip breakers in the closed positten, the CEA drive system capable of CEA withdrawal, and, fuel in the reactor vessel.

1The provisions of Specification 3.0.4 are not applicable.

~4 (a) Trip may be manually bypassed above 10 % of RATED THERMAL POWER; bypass shall' be automatically removed when THERMAL POWER is less.than or equal to 10 '% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 400 psia; bypass shall'be automatically removed whenever pressurizer pressure is greater than.or equal to 400 psia.

~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL P0WER; bypass shall be autcmatically removed when THERMAL POWER is greater than.or equal to 10 *% of RATED THERMAL POWER.

During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be autematically' removed when THERMAL POWER is greater than or equal to 1% of RATED THERu.AL POWER.

(d) Trip may be bypassed during testing pursuant.to Special Test

. Exception 3.10.3.

SeeSpeciaiTestExcepthon3'.10.2.

(e)

(f)

Each channel shall' be comprised of two trip breakers; actual trip logic shall be. one-out-of-two taken twice.

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(g)', Trip.may be bypassed below 55% RATED THERMAL POWER.

ACTION STATEME'NTS With the number of channels OPERABLE one less than required by ACTION 1 the Minimum Channels OPERABLE requirement, restore the h~

inoperable channel 'to OPERABLE status within t.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

With the number of channels OPERABLE one less than the Total ACTION 2 Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed-in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

If the inoperable ' channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with-Specification 6.5.1.6k.

The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN..

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• Jy ivP SAN ONOFRE-UNIT 2 -

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100F & R&i&EOPY ADMINISTRATIVE CONTROLS MEETING FREOUENCY 6.5.1.4 The OSRC shall meet at least once per calendar month and as ' convened by the OSRC Chairman or his designated alternate.

OUORUM 6.5.1.5 The minimum quorum of the OSRC necessary for the performance of the OSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The Onsite Review Committee, shall be res' onsible for:

p a.

Review of 1) all procedures required by Specif.ication 6.8 and changes thereto, 2) all programs required by Specification 6.8 and changes-thereto, 3) any other proposed procedures or changes thereto as

. determined by the Station Manager to affect nuclear safety.

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b.

Review of all proposed tests and exp.erjpents that affect nuclear safety.

c.

Review of all proposed changes.to Append.ix "A" Technical Specifications.

d.

Review of all proposed changes or modifications to unit s'ystems or equipment that affect nuclear safety.

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e.

Investigation of all violations of thi Technical Specifications-including the preparation and forwardi.ng of reports covering-evaluation and r.ecommendations to prevent rec' rrence to the Nuclear u

Control Board (NCB).

f.

Review of events requiring 24-hour written notification to the Commission.

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. Review of unit operations to detect potential nbclear safety hazards.

h Performance of special reviews,. investigations or analyses and l

reports thereon as requested by the Station Manager or the NCB.

l 1.

Review of the Security Plan and implementing procedures and shall j

submit recommended changes to the NCB.

j.

Review of the Emergency Plan and implementing procedures and_'shall submit recommended changes to the NCB.

' k. - Review and documentation of judgment concerning prolonged operation in bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass s,ince the last OSRC meeting.

SAN ONOFRE-UNIT 2 6-7

w-1 ENCLOSURE 2

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' DESIGN CRITERIA In present and various past Combustion Engineering (CE) applications, applicants have proposed to operate four-channel protection systems (Reactor protection System and specific Engineered Safety Feature Systems) with one of the four channels of a given process variable in bypass for an indefinite period of time.

Operating reactor licensees who. desire to use the Technical Specifications of Enclosure 1 must verify that they have re-viewed the design and installation of their protection system and determined that the system meets'the criteria below.

The licensees must also confirm l

that detailed information verifying compliance with the criteria is' avail-able at the licensee's facilities for staff audit.

Until the licensee's confirmation that the p,rotection system meets the criteria below has been completed and submitted to the staff,' bypass of a protection system channei should be limited to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at which time the channel shall be placed in the trip mode.

1.

High Energy Line Preak

-The protection system shsuld be reviewed for the effects of high energy line breaks.

Each licensee must analyze the protection system to verify that high energy'line hazards in coincidence with the bypass of a channel will not negate the minimum acceptable redundancy required by IEEE Std. 279-1971.

It should be noted that credit is not.to be taken for the " fait-safe" mode of the channels affected by high energy line breaks.

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2.

Sinole Failure In Combination With~ prolonced Byoass Th'ere may be cases where the prolonged bypass of a specific protection channel in combination with a single failure might jeopardize plant protection (i.e., channels remaining will not sufficiently detect associated transients and accidents without causing unacceptable consequences such" as core damage, etc.)

The licensee should review the accident analyses (i.e., rod drop accident, rod ejection, etc.) to verify that the bypass of a specific prot.ection channel in coincidence with a single failure of a redundant channel will not prevent re-quired protection for any transient or accident.

3.

Channel Indepenifence The four protection channels must be reviewed for physical in-

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dependence.

Each licensee should confirm,that the four protection channels as installed meet the physical independence criteria of Regulatory Guide 1.75.

4.

Independence of the Vital Buses Each plant must be reviewed for independence of the vital buses.

The Combustion Engineering (CE) reactor protection system (RPS) is made up of foui- (4) protection channels for each trip parameter.

l Each parameter channel consists of bistable relays and associated l

contacts which are arranged into six logic NDs ( AB, AC, AD, BC, BD, '-

CD matricies) which represent all possible coinddences of two com-binations (e.g., combinations of two-out-of-four logic).

'Each logic matrix is powered by two of four Class lE i.ndependent

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120 Vac vital buses as shown in Figure 1.

This arrangement may challenge the isolation and hence independen.ce of_ the redundant ac vital power l

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' buses.

It is typical of licensees using the CE design to assure

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P that the independence of these buses is maintained through th'e use of qualified isolators'.

.a Licensees desiring to use the Technical Specifications of Enclosure 1 should confirm that tests and analyses have been performed to demonstrate independence of the redundant vital buses.

The tests and supporting information should include:

a) The use of a plant-specific mock-up representing one protection logic matrix system (i.e., two matrix power supplies, each with its own simulated 120 Vac vital bus supply, matrix relays, bistable power supplies, bistable tripunits,andisolationcircuitry)..

b) The application of surges (internal and external transient

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voltages) and faults (including continuous phase-to-phase short-circuits, phase-to-ground short-circuits an,d the application of continuous external high voltages) to t'he simulated 120 Vac vital bus supplying power to an associated matrix power supply, c) Application' of the surges and. faults between each matrix power supply input conductor and ground (common mode) and across (line-to-line) the matrix power supply input conductors (transverse riode),

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d) Monitoring the redundant simulated 120 Vac vital bus supplying power to its matrix power supply'to measure any effect as a re-sult of application of the faults or surges on the other bus, c

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formed.

Licensees desiring to use the Technical Specifications of Enclosure

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1 should confirm that sufficient tests and analyses have been per-formed to assure that with a channel bypassed, a ~ vital bus single failure will not negate the required prote.ctive function.

The tests and supporting information should include:

a) The use of a plant-specific mock-up representing one pro-tection logic matrix system (i..e., two matrix power supplies, each with its own simulated 120 Vac. vital bus supply, matrix-relays, bistable power supplies, bistable trip units, and isolation circu,itry),

b)

The application o,f surges (internal and external transient voltagss) and faults -(including continuous phase-to-phase short circuits, phase-to-ground short-circutts and the application.of continuous external high voltages) to tne simulated 120 Vac vital bus supplying power to an associated matrix power supply, cl. The application of surges and faults between each matrix power supply input conductor and ground (common mode) and across (line-to-line) the matrix power supply input con-ductors (transverse mode),

d) Moni.toring the auctioneered matrix. power supply output to measure i

any effect on the logic matrix circuitry as 'a resul't of application.

of the faults or surges, 1*

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e) Verification that during and after the application of the surges and faults, the protection circuits will perform their protective actions, f) Justification ~ that the faults and surges used 'during the testing exceed the maximum worst-case failures which could occur within the protection systems circuits.

Any response addressing the degree to which the ANO-2. design meets these criteria may reference information previously submitted on the ANO-2 docket on these subjects.

The reference should be specific with respect to the location of the specific in-formation and should state the bases for the licensees conclusion that the infor-mation supports the conclusion that the ANO-2 design meets the subject criteria.

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