ML20084S377

RO 50-254.76-34-on 761104,electromatic Relief Valves 1-203-3C & 1-203-3E Failed to Open When Actuated from Control Room.Caused by Excessive Steam Leakage Into Area Below Valve Disc.Investigation in Progress W/Manufacturer

ML20084S377
Person / Time
Site:	Quad Cities
Issue date:	11/12/1976
From:	Kalivianakis N, Schrock J COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NJK-76-420, RO-50-254-76-34, NUDOCS 8306160709
Download: ML20084S377 (4)

Text

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CoMm'onw2:lth Edison Ouad-Cities Nuclear Power Station b IE FILE COPY Post Office Box 216 , ,. .

Cordova, Ilknois 61242 < , -^ k',/ M Telephone 309/654-2241 NJ K-76-420 A 6 g + 'e Rittat0 November 12, 1976 m2 DEC 71976 d a -

"d " Ess w semes 5

J. Keppler, Regional Director /

Office of Inspection and Enforcement 11 TO\

Region ill U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137

Reference:

Quad-Cities Nuclear Power Station Docket No. 50-254, DPR-29, Unit 1 Appendix A, Sections 3 5.D.3 and 6.6.B.I.e.

Enclosed please find Reportable Occurrence Report No. Ro 50-254/76-34 for Quad-Cities Nuc, lear Power Station. This occurrence was previ.ously reported to Region ill, Of fice of Inspection and Enforcement by' telephone on November l',

1976 and by telecopy on-November 2, 1976.

This report is submitted to you in accordance with the requirements of Technical Specification 6.6.B.I.

Very truly yours, COMMONWEALTH EDIS0N COMPANY QUAD-CITIES NUCLEAR POWER STATION

' /

W N.J. Kallvianakis Station Superintendent

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NJK/LFG/lk cc: G.A. Abrel1 lb c 2 8306160709 761112 PDR ADOCK 05000 S

NOV 101976

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i. LICENSEE EVENT R RT CONTROL 8 LOCK:l l l l l l l (PLEASE PRINT ALL REQUIRED INFORMATION) 1 6 AME LCENSE NUMBER E TYPE O1 ll lL lQl Al Dl 1l l0l0l-l0l0l0l0l0l_l0l0l l 4l 1 l1l 1l 1l l0l1l l 7 89 14 15 25 26 30 31 32 CATEGCRY TYP E OCCKET NUMBER EVENT DATE REPORT DATE U59 l1 l1 l1 l2 l 7 l 6[

O CONTl l l l Ll l0 l 5 l 0 l-l 0 l 2 l 5l 4l l1 l110 l 1 l 7 l 6l 7 8 57 58 60 61 68 69 74 75 80 EVENT DESCRIPTION gg l During the startup for Unit Two, Cycle Three, the electromatic relief valves were l 7 89 80 SE l tested for operability. Two of the relief valves failed to open. Due to these l 7 89 80 g[ l problems on Unit Two, it was decided to test the Unit One electromatic relief valves. [

7 89 80 3E l At 2:45 pm on November 1, 1976, the Unit One relief valves were tested. The 1-203-3Cl 7 89 80 HG I and 1-203-3E relief valves failed to open when actuated from the control room. it wasl 7 89 pauf 80 E ODE COMPONENT CODE V101ATION 35 l Sl Fl W 7 89 10 11 l V l Al Ll Vl ElXl 12 17 W43 l Dl2 l4 l3 l 44 47

{48 CAUSE DESCRIPTION gg l (Proximate Cause-Equipment Failure) These valves are designed to operate by venting l 7 89 80

@ l the area below the valve disc by opening a pilot valve, and causing a differential J 7 89 80 DE I pressure across th4 valve disc. This results in the valve being forced- (see attached)

S  % POWER CTHER STATUS DS VERY DISCOVERY DESCRIPTON 1

7 8 W

9 l0l7l7l l 10 12 13 NA 44 l W45 l Test based on Unit 2 valve failures l 46 80 R LEAS D OF AMOUNT OF ACTIVITY LOCATON OF RELEASE 30 7 8 W

9 dl 10 11 NA 44 l l 45 NA 80 l

PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTON

(( l 0l 0 l 0 l 7 89 11

]12 l

13 NA 80 l

PERSONNEL INJURIES NUMBER DESCRIPTON 3E l 0l 0 l 0 l l NA l 7 89 11 12 80 OFFSITE CONSEQUENCES

@l 7 89 NA 80 l

LOSS OR OAMAGE TO FAClUTY TYPE DESCRIPTON .

32 U 7 89 l

10 NA 80 l

PUBLICITY DE l NA l 7 89 80 ADDITIONAL FACTORS gg l (Event Description contd) determined that the pilot valves were functioning as l 7 89 80 i DE l evidenced by a temperature rise down stream of the pilot valve. However, (cont) l 7 89 80 NAME: James Schrock PHONE:

GPO 801

(3 V (v)

CAUSE DESCRIPTION contd open by reactor steam pressure. There is an orifice through the disc retainer that allows the area below the valve disc to re pressurize and close the valve when the pilot valve is closed. The apparent cause of the valve failures was that there was excessive steam leakage into the area below the valve disc, in addition to the normal steam flow to this area through the orfice. As a result of this additional steam flow, the pilot valve could not adequately vent the area below the valve disc to allow the valve to open.

There are two possible leakage paths that could have resulted in the valve failures. The steam could have leaked between the valve disc guide and the piston rings on the valve disc or there could be leakage past the threads on the disc retainer.

ADDITIONAL FACTORS EVENT DESCRIPTION contd as described below, there was no noticeable steam flow from the reactor through the valve discharge piping.

The presence of steam flow was checked by first lowering the setpoint on the turbine Load Set, in order to open a main steam bypass valve. Each relief valve =was actuated from the control room for a one-second time interval.

Relief valve opening and subsequent steam flow are then verified by means of observing closure of the bypass valve during the time the relief valve is.

open.

This response of the bypass system was satisfactorily verified for relief valves 1-203-3A, 1-203-38, and 1-203-3D.

In accordance with Technical Specification 3.5.D.3, an orderly shutdown was commenced immediately and load was decreased at the rate of 50 MWe per hour.

The unit was in the Cold Shutdown Condition by 8:00am on November 2, 1976.

Work Requests 4094-76 and 4095-76 were issued to determine the source of the problem and perform repairs.

The failure of two electromatic relief valves renders the Auto-blowdown function of the Emergency Core Cooling System inoperable. However, the High Pressure Coolant injection (HPCI) Sub-system was demonstrated to be fully operable, thereby providing a means available for introducing emergency core cooling water into the reactor vessel at operating pressure. Ancther function of the electromatic relief valves is to protect the vessel from over pressurization. This function was provided for by virtue of the three operable relief valves providing a path for blowdown to the suppression chamber. The Target Rock Safety-Relief Valve 1-203-3A was demonstrated to be operable, and would have lifted at a reactor pressure of 1125 psig.

Valves 1-203-3B and 1-203-3D would have opened at reactor pressures of 1130 psig and 1135 psig, respectively. Therefore, there was overpressure protection at all three pressure setpoints specified in Technical Speci-fication 4.6.E. Also, the Electro-hydraulic Control (EHC) system was operable,

enabling the bypass valves to dump steam to the condenser in the event of a turbine trip, thereby controlling reactor pressure. Furthermore, all i nine Main Steam Safety Valves were operable, and were fully capable of preventing the reactor vessel pressure from reaching the Safety Limit of 1325 psig at the vessel steam space. Therefore, the possible consequences of this occurrence were minimized by the redundant design of the safety systems and the fact that all other safety systems were operable. At no time was the public health and safety in Jeopardy, nor was the ability to safely shutdown the reactor compromised. (R0 50-254/76-34)

Corrective Action to Prevent Recurrence When Unit One was shut down and de-inerted, a drywell entry was made. The two elecromatic relief valves that failed were removed, brought to the shop, disassembled, and inspected. The disc retainer lock arm and lock screw was missing from'the 1-203-3E relief valve. It was decided to replace the 1-203-3E valve, serial number 7069, with a spare relief valve, serial number 7062, and finish inspecting the 7069 valve at a later date. The spare relief valve had been overhauled and had new piston rings, valve disc, and pilot valve disc installed. The 1-203-3C relief valve, serial number 7063, was overhauled and the piston rings and valve disc guide were replaced. Both valves were then given a leak test and operability test in the shop and then were replaced on the main steam lines. Startup was then commenced on Unit One on November 6, 1976. The reactor was brought to operating pressure and the 1-203-3C and 1-203-3E relief valves were tested using Temporary Procedure No. 743 This procedure constitutes a revis. ion to the existing procedure for manual operation of the electromatic relief valves, .and calls for verification of a bypass valve closure response as well as temperature, to verify steam flow from a relief valve. This revision will be a permanent procedure, and shall be implemented accordingly. After the electromatic reliefs were replaced on the Unit One steam lines the 7069 relief inspection was completed and the threads on the disc retainer were found to be worn.

Investigations into this problem are being continued between the valve manufacturer and Commonwealth Edison Company to resolve the causes of these valve failures.

Failure Data Unit One Main Steam Relief Valves had not experienced this type of failure in the past. Modification M-4-1-73-45 was installed on the relief valves in May, 1973 This modification installed a valve disc retainer locking device, to prevent the locking plate from falling off. This change had been performed ba::ed on an electromatic relief valve failure at Oyster Creek Station in December, 1972.

The 1-203-3C and 1-203-3E electromatic relief valves are manufactured by Dresser Corporation, serial numbers BK-7063 and BK-7069

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