ML19224B859

Summary of 790509 Meeting W/Util to Evaluate Emergency Feedwater Sys in Terms of TMI Incident

ML19224B859
Person / Time
Site:	Crane, Arkansas Nuclear
Issue date:	05/15/1979
From:	Engle L Office of Nuclear Reactor Regulation
To:	Boyd R, Case E, Ross D Office of Nuclear Reactor Regulation
References
NUDOCS 7906270097
Download: ML19224B859 (13)

Text

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NUCLE AR REGULATORY CCMMISSION f g -.,,f),. h now:a rem. o. c. wsss f~

e fO l Docket No:

50-368 LICENSEE: Arkansas Power & Light Company FACILITY: Arkansas Nuclear One, Unit 2 (ANO-2)

SUMMARY

OF MEETING FOR ARKANSAS NUCLEAR ONE, UNIT 2 (AN0-2)

REGARDING THE EMERGENCY FEEDWATER SYSTEM A meeting was held on May 9,1979, regarding the subject as noted above. The purpose of the meeting was to evaluate the Emergency Feedwater System (EFS) for ANO-2 in light of the Three Mile Island incident.

A list of attendees is provided in Enclosure 1, and a list of questions sent to the licensee by the staff on May 4, 1979 by way of telecopy is provided in Enclosure 2.

The licensee stated that the EFS for ANO-2 is designed to automatically initiate as part of the Engineered Safety Features Actuation System.

The system is designed to the latest revision (Rei.1) of Branch Technical Position ASB 10-1.

The EFS is designed to provide a means of supplying water to the intact steam generator (s) following a postulated main steam line rupture or loss of main feedwater to remove reactor decay heat and provide cooldown of the Reactor Coolent System to the temperature and pressure at which the Shutdown Cooling System can t;e placed in operation.

Redundancy is provided for components of the EFS to assure operation in the avent of a single failure of a mechanical or electrical ccmponent within the system.

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The EFS system employs one_lurbine_drivertpum_p, one_niotor driven pump, and two independent feedwater trains each capable of supplying eftEer dT tire two steam generators.

The pumps and piping system, except for the, condensate water supply and flush and recirculation lines downstream of the firstlso'lation valves are designed to meet ASME Section III, Class 3 and Seismic Category I requirerrents.

The isolation valves are designed to meet ASME Section III, Class 2 and meet Seismic Category I requirements.

At rated flow, each pump is capable of providing sufficient makeup water to the steam generators for removing a decay heat load of 3.5 oercent of full reactcr power at maximum steam generator pressure.

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' The EFS discharge piping and valving arrangement is designed to allow either pump to supply cooling water to either or both generators.

Each scpply line to each steam generator is provided with redundant control valves in accordance with single failure criteria to ensure isolation of the steam generators and feeding of the remaining intact steam generator as required during an enaineered safety features actuation of the EFS follcwing a postulated main steam or feedwater li ie break.

ihe EFS is provided with the necessary controls for local or remote and automatic or manual operation of the system. Local controls are mounted on a panel near the pumps and the remote controls are in the main control room.

All controls and control signals for the steam turbine - driven pump and the electric motor driven pump are channelized.

Physical separation between the electrical components is provided in accordance with IEEE 279-1971 Standard Criteria for Protection Systems for Nuclear Power Generating Stations.

The desicn basis events which will cause autcmatic emergency feedwater actuation (1} steamline break inside containment, (2) steamline break outside containment are:

dnd (3) loss of main feedwater.

The parameters which are monitored to indicate these events are (1) steam generator pressure and (2) steam generttor level.

The staff itemized in detail the licensees response to the questions provided in and indicated to the licensee that the results of the staff's evaluation on the ESF for ANO-2 would be provided in a document to be issued in the near future.

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'A p MJ Leon B. Engle, Pr ject Manager Light Water Reactors Branch No. 1 Division of Project Management

Enclosures:

1.

Attendance List 2.

Request for Information, dated May 4, 1979 cc:

See next page

' r, c c/g no jusJ

Mr. William Cavanaugh, 111 Mr. William Cavanaugh, III Executive Director of Generation & Construction Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203 Mr. David C. Trimble cc:

Manager, Licensing Arkansas Power & Li it Company P. O. Box 551 Little Rock, Arkansas 72203 Philip K. Lyon, Esq.

House, Holmes & Jewell 1550 Tower Building Little Rock, Arkansas 72201 Mr. E. H. Smith, Project Engineer Bechtel Power Corporation San Francisco, California 94119 Mr. Fred Sernatinger, Project Manager Combustion Engineering, Inc.

1000 orospect Hill Road Windsce, Connecticut 06095 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations C-E Power Systems Conbustion Engineering, Inc.

4853 Cordell Avenue, Suite A-1 Bethesda, Maryland 20014 Mr. James F. O'Hanlon General Manager - Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 255

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ENCLOSURE 1 ATTENDANCE LIST FOR MEETING, MAY 9, 1979 ARKANSAS NUCLEAR ONE, UNIT 2 ARKANSAS POWER & LIGHT COMPANY AP&LCo.

Sandia Corporation P. L. Almond G. Kolb B. A. Baker NRC Staff R. Cook L. 8. Engle J. T. Enos J. Calvo E. Ewing M. Greenberg G. Young

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ENCLOSURE 2, MAY 4, 1979 REQUEST FOR ADDITIONAL INFORMATICN As part of*its on-going review of the Three Mile Island Unit 2 a:cicent, the staff finds that it needs additional information regarcing t.e auxiliary feedwater systems (AFWS).

This information as outlined below, is rcquired to evaluate AFWS reliability for Combustion Engineering (CE) and Wes:ing at;e designed pressurized water reactors.

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requested information is in additior to that requested in the IE Bulletins, t.nd should be brought to the reeting scheduled with the staf f on May 8 thru 4ay 12,1979.

Written system description (as built) including:

- List of Support Systans for Auxiliary Feed System Operation (Soth Electric and Steam)

- Water Supplies for AFWS (primary and Lc-Lup)

Current operating procedures and test and maintenance requirements including:

- All LCO's for AFWS, main FW system and related support system 3.

- Listing of operator actions (local and/or control rocm) and timing require-ment; for such actions.

Procedures for reinitiating rnain feedwater flow.

As Built P& ids with symbol keys including condensate and steam side Ledgible Equipment layouts drawings including:

Isometrics, if available Identification of inhibits preventing accessibility to AFWS components and related electrical equipment Relevant control systems description including:

- Schematic or logic control diagrams

- Listing of actuation signals / logic and control MSIS logic for isolating AFWS, if installed

- electric power dependences

- All " readouts" available in control room for AFWS operation AC & DC Power

- One line diagrams (normal and emergency pcwer supplies)

- Divisional designation e.g., Train A, Trai" B, requirements on all AFWS components end support systems

- List of normal valve states and loss-of-actuation pcser failure position Operating Experience, including

- Number of nain feedwater interruptions per year experienced te date for each unit

- Number of demands on AFWS per year to date (test and actual) for each unit

- Summary of AFWS mal functions, problems, failures Provide Available reliability analyses Steam Generator dry-out times (assuming loss of all f eedwater flow, with 10C0 initial pcwer, with Reactor trip, no line breaks)

System design bases including:

- Seismic and environmental qualification Code and Quality, QA

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Faae c Ma;. 4, 1979 Provide written responses to the following set of cuestions by 5/8/79

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Describe backup sys tems available (to auxil:ary feedaater) for pro. iding teedwater to stear g e r.e ra to rs.

Discuss actions and time re:uired to make these systems available. Are procedures available?

If so, provide.

Provide the folicwing procedures:

loss of offsite power loss of feedwater LOCA (small and large)

Steam Line Break Provide following information for PORV's:

Number capaci ty setpoints (open and close) manufacturer and model

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indications of position record of periods isolated (isolation valve shut) challenges during life of plant (from plant records) including performance of valve, cause of challenge.

experience of two-phase or subcooled discharge of PCRVs and safety valves with description of valve performance Provide indications of PORV isolation valve in the control room.

Provide the following information on ECCS:

initiation setpoints system description pump performance characteristics (head curves)

Provide reactor protection system trip setpoints.

Provide information on charging pumps, how they relate to ECCS including:

numbe-ficw vs. pressure power sources and backup water sources seismic qualification List all challenges (and cause) to ECCS as indicated on plant records.

List and discuss all instances during which your plant has undergone natural circulation.

Describe all automatic and manual features which can stc; tne reactor coolant pumps.

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i t i me h i At'o ry o f t he t a l l md -ise at *he w a :.; r 1._.!

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As stated at the ACRS meeting in Russellville, li r, Eh?rsul's e r,c e rn s s.:emed to he with small break si:es ".

. up to ind. inst bey on.J the capacity ot the cha rging pumps, i

LO-2 has three (3) cha rging ptuaps each with a iapacite 4 approxinat'ly

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t 13 r,pm for a total of 12S Thi s flow ra t e is approximate' equal to

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p a break si:e of 0.003 f t2,gpm.The ECC3 Smal1 Ibeaks Ana1ysis submitted to U

NRC ca September 30, 1977 stated, " Calculations indicate that break sizes cit approximatel[U.Ui it: and smaller will not un M r v roit

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gi TherCfDirty this-anat) si s there-is noTohililii ty a F.m.meric dmre

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with break si:es "up to and just beyer.d" the cap.icity of the charging

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ya filowever, to adu res:> the question core specificall;., for a rapt u re twice the the capacity of the cha r'ging pwnps, operator.ict ion wonid connence as

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shown in Operating Procedure 2202.06 page 1 and 2 (attached).

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'. an init ial pres suri zer level indicat ion of 50"., a break of this magnitude j ' ) e/

would Icwer pressuri:er level at the rate of 33 indication per minute therefore allowing over 16 minutes for the eperators to respond before p

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0~. pressari:er 1cvel was reached.

As per procedure, the inability to h.a

.g maintain pressuri:cr level with the charging pumps would result in a

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.f y p T.it oi i ni t ia t H rnS System depressuri::..oa would then be tu im a hately to.i p rt. ssu re be l ow the !! PSI pu:np max imum hoad.

~ihi., depressuri_ation

[g. g could be almost instanevus without vietating technical.peci fic:it ions v.d w i t hout initiating cooldown.

Therefore, even though the core would not be uncovered i f there were no operator action, it is our conclusion Operating Procedure 2202.C0 assures that t he pl:.nt can be depressed and brought to a lang tern coo 1in" con-dition fo. a small break beyond the c ipaci ty of the chart n; pumpr.

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