ML20085M729

Interim Deficiency Rept Re Design Deficiency in Control Logic of Reactor Bldg Hpci/Rcic Sys.Initially Reported on 731110.Design Changes Proceeding.Final Rept Will Be Submitted by 740110

ML20085M729
Person / Time
Site:	Browns Ferry, 05000000
Issue date:	12/14/1973
From:	Gilleland J TENNESSEE VALLEY AUTHORITY
To:	Kruesi F US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20085M697	List: ML20085M723 ML20085M729 ML20085M744 ML20085M751
References
10CFR-050.55E, 10CFR-50.55E, NUDOCS 8311090186
Download: ML20085M729 (5)

Text

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'* f TENNESSEE VAL '.Y AUTHO AITY E"'"3M CHATTANDOu , TENNESSEE I. O 6 37401 L '

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December 114 , 1973 $$5N[jf Mr. F. E. Kruesi, Director. '

Directorate of Regulatory Operations 1 U.S. Atomic Energy Commission .

l Washington, DC 20545 i

Dear Mr. Kruesi:

TVA made initial report to AEC-DRO Region II on November 10, 1973,

, of a possible design deficiency in the control logic of HPCI and RCIC systems of our Browns Ferry Nuclear Plant. An interim report

'on this deficiency was submitted to you on December 10, 1973. En-closed is a second interim report on the safety implications associated with this problem. We will submit a final report by January 10, 1974, on design changes being made.

Ve truly yours, fl'5h&Gd J'. E. Gilleland Yf Assistant to the Manager of Power Enclosure CC (Enclosure):

Mr. Norman C. Moseley, Director Directorate of Regulatory Operations

~

U.S. Atomic Energy Commission i

Region II'- Suite 818

• 230 Peachtree Street, NW.

l Atlanta, Georgia 30303 4

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ENCIASURE BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 HPCI AND RCIC CONTROL LOGIC PROBLEM INTERIM REPORT ,

l

References:

(1) Letter from E. F. Thomas to John F. O' Leary, AEC, Director, the Directorate of Licensing, Office of Regulation, " Tennessee I

Valley Authority - Browns Ferry Nuclear Plant Unit 1 - Docket No. 50-259 - Facility Operating License DPR Abnormal Occurrence Report BFAO - 7334W," with enclosure, dated November 19 -1973.

(2) Letter from R. B. Beers to H. M. Bankus, General Electric Company Internal Letter No. 4668, " Pipe Break Outside Contain-ment: Main Steam Line Break, No RCIC, No HPCI, No foss of Offsite Power," with attachment, dated June 1, 1973. Copy is attached.

He initial report of the HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) control logic problem of units 2 and'3 at the Browns Ferry Nuclear Plant was made on November 10, 1973, to the AEC Directorate of Regulatory Operations, Region II Office in compliance with paragraph 50.55(e) ,

of 10CFR50. ,

Under the operating situation described in reference 1 for the unit i reactor, the momentary loss of offsite power would trip the Fee'dwater Control. System-

-(FCS) as well as the HPCI and RCIC systems. The operator would have been unable to start the RCIC and HPCI systems until the standby. diesel generators energized the shutdown auxiliary. boards and the tripping logic to the RCIC and HPCI systems was manually reset. Since units 2 and.3 of the plant are designed similarly to unit-1, we assume that similar occurrences.can later

~

take place when unita 2 and 3 are operated unless appropriate changes are made.-

The safety implications associated with this problem would be no worse than -

those produced during the situation described in reference 2l(copy attached) even if the plant had been at fullipower-and had contained an equilibrium inventory of fission products.x The= situation in reference 2 assumed a main f

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1 steam line break'with no RCIC, no HPCI, and.NE loss of offsite power. A main steam line break causes closure of the main steam line isolation valveo; a loss of offsite power causes isolation of the steam lines by either rapid closure of the steem turbine control valves or by closure of the turbine stop valves. There would be no steam dump with no offsite power. Thus, either situation isolates the steam lines and requires steam relief to the torus. The curves in reference 2 show that the reactor is safe for 10 minutes with no steam flow except through the relief valves and no RCIC or HPCI. The presence or lack of offsite power during this time int'erval would not affect

. the results. Note in reference 2 that after 10 minutes the Automatic Depressuri-zation nystem (ADS) was actuated by operator action. It could also be assumed that, had it been necessary during the present problem, the ADS could have been manually actuated after 10 minutes. Also note in the curves in reference 2 that the vessel pressure did not exceed 1080 psia, the peak clad temperature did not exceed 1200*F, and an adequate vessel water level was maintained.

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ATOMIC PO'.lER EQUIP,'EilT LEPART 'EilT - -

San Jose, California Internal Letter !!o. 4658 cc: RR Barris JE Stice June 1, 1973 BN Rogers H. I1. Danims Knoxville Office 14AIll STEAL'LIi!E BREAK, H0 RCIC,

Subject:

PIPE BREAK OUTSIDE C0HTAltlMEiIT:

N0 llPCI, U0 LOSS OF 0FFSITE POWER Recently TVA (Jerry Chapman) has postulated a steamline break in the steam tunnel which disa' oles the RCIC. They also assumed a single failure of the

. HPCI tr;th offsite power available and drywell coolers operating.

Design Engineering lias reviewed the case of a steamline break outside the ,

containment with no RCIC, no HPCI, no loss of offsite power and drywell coolers operating for a 251-764 '67 product line plant. They have concluded that the peak clad temperature is well below the perforation threshold and -

no fuel rod failures will occur. Adequate water level is maintained and ,

10 minutes following the accident the operator is assumed to take acticn by -

manually activating the ADS system.

Please informally advise TVA (Jerry Chapman) of this fact, f,

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Acting llanager .

Browns Ferry Project .- .

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