ML19322D924

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Current Events for Sept-Oct 1977
ML19322D924
Person / Time
Issue date: 12/31/1977
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NRC COMMISSION (OCM)
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ML19322D920 List:
References
PR-771231-02, NUDOCS 8003110210
Download: ML19322D924 (10)
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REGULATORY POWER REACTORS commission THIS COMPILATION OF SELECTED EVENTS IS PREPARED TO DISSEMINATE INFORMATION ON. OPERATING EXPERIENCE AT NUCLEAR POWER PLANTS IN A TIMELY MANNER AND AS OF A FIXED DATE. THESE EVENTS ARE SELECTED FROM PUBLIC INFOPM TION SOURCES.

NRC HAS, OR IS TAXING CONTINUOUS ACTION ON THESE ISSUES AS APPLICABLE, FROM AN INSPECTION AND ENFORCEMENT, LICENSING AND GENERIC REVIEW STANDPOINT.

1.

1SETBER - 31 OCTOBER 1977 (PL'3LISHED DECEMBER 1977)

OPERATOR ERROR On January 11, 1977 while the Fort Calhoun Station Unit 1 was operating, water from the Refueling Water Storage Tank was pumped into the containment through the containment spray header due to an operator error.

During the perforr.ance of a quarterly test of the safety injection b

and containment spray pumps, the operator noticed an increase in h

f the containmer.t sunp level approximately ten minutes after the low pressure safety ir.jection pump had been started.

Approximately l

3300 gallons cf water had been pumped to the containrc:ent.

About one r

minute later the ventilation isolation actuation signal was received.

i-At this time the cperater realized he had failed to follow the sur-veillance procedures and had left the discharge valve of the low head i

safety injection pump open.

He iranediately secured the pump.

The Reactor Ccolar.t System was checked for leakage and containment entry was made approximately one hour later.

Inspection revealed that a discharge from the containment spray nozzles had occurred.

A few minutes later pcwer reduction was started.

A second containment entry was made about an hcur later, after containment air samples confirmed that a full face mask wodid provide adequate respiratory protection for the levels of radioactivity in the building.

A detailed inspecticn revealed no serious deficiencies and no electrical grounds; the power reduction was terminated at a power level of 83%.

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Although tna cperator had not followed the procedure and the discharge i

valve was open, the centainraent spray header isolation valve (HCV-345) h."

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and the low pressure safety injection to containment spray header

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cross-connect valve (HCV-335) should have prevented the event.

The electric / pneumatic converter on HCV-345 had failed and both red and green pcsition indication lights were on, indicating the valve was partially open. Prior to the event the auxiliary Building Equipment Operator had'taken local control of the valve in an attempt to com;1etely close the valve. After about 1/2 inch of stem travel, the operator re=oved the valve pin and the valve went back to its previous position as demanded by the valve positioner.

The third valve (HCV-335) in the incident had a leakage problem that had been previously identified but no torrective action had been taken.

The pneumatic relay on valve HCV-345 was replaced and valve HCV-335 repaired.

Yalve HCY-344 and HCV-345 are now required to be placed in the test mode prior to operating the low pressure safety injection pump or contain spray ptmp for testing.

This mode along with verifi-cation of an annunciator will ensure that both of thepe valves are in the fully closed position prior to pump operation.'

VALVE MALFUNCTIONS si:.

1.

Prirary System Depressurization On September 24, 1977, Davis Besse Nuclear Power Station Unit e

P No.1 experienced a depressurization when a pressurizer power relief valve failed in the open position.

The Reactor Coolant Systea (RCS) pressure was reduced from 2255 psig to 875 psig io approxicately twenty-one (21) minutes.

At the beginning of f;.-

this event, steam was being bypassed to the condenser and the reactor thermal power was at 263 MW, or 9.5%.

Electricity was not being generated. The following systems malfunctioned during the transient:

a.

Steam and Feedwater Rupture Control System (SFRCS).

b.

Pressurizer Pilot Actuated Relief Valve.

c.

No. 2 Steam Generator Auxiliary Feed Pump Turbine Governor.

The event was initiated at 2134 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.11987e-4 months <br />, when a spurious " half-trip" occurred in the SFRCS, resulting in closure of the No. 2 Feedwater Startup Valve and loss of flow to No. 2 Steam Generator.

Approxi-mately one minute,later, low level in the No. 2 Steam Generator caused a full SFRCS trip, closing the Main Steam Isolation Valves

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(HSIV). The loss of heat sink for the reactor caused the RCS temperature, pressure, and pressurizer level to rise.

The RCS pressure increased to the pilot actuated relief valve setpoint (2255 5sig) and the valve cycled open and closed nine tines in rapid succession, failing to close on the tenth opening.

Meanwhile, the reactor operator observed the pressurizer level increase and manually tripped the reactor about one minute after MSIV closure (two minutes into the transient).

At this point the RCS pressure was approximately 2000 psig and decreasing while the pressurizer level had reached its maximum initial rise of about 310 inches. The RCS pressure continued to decrease due to the open relief valve and upon reaching 1620 psig approxi-4 mately three minutes into the transient, actuated Safety Features

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including high. pressure (water) injection and containment isolation.

Approximately five minutes into the transient the rupture disc on the pressurizer quench tank, which was receiving the RCS blowdown, burst.

Bursting of the rupture disc was aggravated by the actuation of containment isolation, which had isolated the quench tank cooling system, resulting in expedited pressuri-zation of the quench tank.

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The RCS contitued to blow down through the open pressurizer power relief valve and the quench tank rupture disc opening until primary coolant saturation pressure was reached, about se six minutes into the transient.

The formation of steam in the RCS caused an insurge of water inta the pressurizer.

This insurge and the high pressure water injection then restored pressurizer level to about 310 inches after nine minutes into the transient.

Approximately thirteen minutes into the transient, the secondary side of the No. 2 Steam Generator went dry.

About fourteen minutes into the transient, the operators noticed the low level condition and found that the auxiliary feed pump was operating at reduced speed.

Manual control of the auxiliary feed pump was started and water level restored to the No. 2 Steam Generator.

At approxirately 21. minutes into the transient, 'the operators discovered that the pressurizer power relief valve was stuck open.

Blowdown via this valve was stopped by closing the block valve, thus terminating the reactor vessel depressurization.

The RCS pressure recove, red to normal and cooldown of the system followed.

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The reason for the spurious " half-trip" of t.he SFRCS has not yet been determined.

An extensive investigation revealed several loose connections at terminal boards, but nothing conclusive.

Investigation into the failure of the pressurizer pilot actuated b

relief valve revealed that a "close" relay was missing from the

.n control circuit. This missing relay would nomally provide a

" seal-in" circuit which would hold the valve open until the pressure dropped to 2205 psig.' Without the relay the power relief valve cycled open and closed each time the pressure of the RCS went above or below 2255 psig.

The rapid cycling of the valve caused a failure of the pilot valve stem, and this failure caused the power relief valve to remain.open.

It was determined that the auxiliary feed pump did not go to full speed because of " binding" in the turbine governor.

The transient was analyzed by the NSSS vendor and determined to be within the design parameters analyzed for a rapid depressurization.

With excepticn of the above noted. malfunctions, the plant functioned as desicned and there was no threat to the health and safety of the

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2.

Feedwater Isolation Valves e

On two occasions in July, at the Trojan nuclear plant, a hydraulic

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feedwater isolation valve failed to close-upon receipt of a close signal.

All other equipment required to operate, functioned

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The first failure, July 6,1977, had been attributed to an

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improperly assembled solenoid in the hydraulic actuator.

Investigation of the second failure indicated that both events were due to a lack of sufficient hydraulic pressure.

Failure of the valve to close was caused by the pressure regulator leaking and failing to close down to regulate the pressure.

This caused the hydraulic system on the valve to be drained down to a point that the valve would not operate.

Inspection of the regulator revealed'that a locking screw on the r.egulator adjusting

' I knob was loose and would allow the knob to vibrate to any position.

With the regulator improperly set it would not close down to regulate pressure and would allow the hydraulic fluid to drain before the hydraulic operator could function.

A similar problem

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All of the regulators were reset and the adjusting knobs were

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isolation v ves were tested satisfactorily following these adjustments 3.

Off-Gas.Systea Yalves At the Oyster Creek nuclear generating station on August 27, 1977, the reacter building ventilation system isolated and the standby gas treatraent system (SGTS) automatically initiated.

j Investigatien revealed that at approximate employee perforrsing housekeeping duties in,1y' 1850 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.03925e-4 months <br /> a station the main control room accidently caused the augmented ' ff gas (A0G) mode switch to move o

fron " isolate and bypass" to the " isolate" position.

This resulted in the cff gas valve and the off gas drain valve going closed, and since the A03 was not in service the gas flow was stopped.

The isolation of the reactor building ventilation system and initiation of the SGTS occurred at 1905.

The two off gas valves were opened four miro%s later and the SGTS was secured.

The reactor building ventildion system was returned to normal at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.

en The off gas drain valve did not. seat properly and was not leak

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This conditien allowed the gaseous radioactivity within the isolated off gas system piping to travel up through the stack

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sum; in the stack base and fill the air space in the ventilation

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r tunnel.

k' hen the raciation level in the reactor building ventilation duct reached a level of 17 mr/hr the monitors located next to this duc-initiated the SGTS.

4 The safety concern associated with this event is the possibility of a sub.,ergence dose a person would have received from the radio-

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active gaseo::s at:r.osphere if they were in the tunnel area.

The atmcsphere ir. the tur.nel area is processed through the radwaste ventilation system, which contains both roughing and absolute filters, prict to exhausting through to the stack which is monitored.

he taxicum radiation level sensed in the tunnel was 26 tr/hr.

4 No persennel exposures or releases to the environment resulted from this event.

The licensee is investigating the feasibility

.7 of installing an alar:a to alert operations personnel.to the closure of the off gas valve when the A0G is out-of-service.3

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Vss SMALL PIPE BREAK ANALYSIS On June 9,1977, an orderly shutdown of the Yankee Nuclear Power Station (Yankee Rowe), a pressurized water reactor, was initiated by the licensee because of an error discovered in the Emergency Core Cooling System (ECCS) performance analysis.

Yankee Atomic Electric Company (NRC) that an error had been discovered (YAEC), the licensee, notified the fiuclear Regulatory Comnission in a particular small break loss of coolant accident (LOCA) analysis, which permitted reactor operation with Core XII. in a manner less conservative than assumed in the original analysis.

While performing a review of the analyzed small break accidents for the Core XIII reload, the YAEC Safety Analysis Group determined on June 7,1977 that an incorrect fluid flow resistance calculation was cade in the safety injection line break analysis.

The fluid flow characteristics study had taken credit for the 2-1/4 inch safety injection line thermal sleeve to retard spillage from the accumulator --

a tank which suoplies borated water to the reactor core in the event cf a reactor coolant system pipe break.

The flow resistance of the sleeve should not have been included in the flow calculation, as a new worst case pipe break was identified in a 4-inch diameter line section.

The recomputed de~ reased flow resistance allowed increased accumulator c

or flow to be calculated for the break, and decreased the ECCS supply pressure to less than had been assumed, thus decreasing the core reficod capability of the ECCS.

This corrected flow resistance assumption was used for the accident analysis of the present core, Core XII, which was operating at 79% of rated power in a coastdown program prior to the June 9, 1977 shutdown.

Operation of the reactor with Core XII commenced in December 1975.

Upon discovering the error, the licensee reduced power level to 300 cegawatts thermal (50% rated power), which was believed to conservatively acconcodate the analysis error.

During subsequent analysis, however, the licensee was unable to assure himself that the 10 CFR 50.46 limits on peak fuel cladding temperature could be raintained for the postulated s all break.

Therefore,.the facility was shutdown pending resolution of this matter and to proceed with the Core XIII refueling outage which had been previously scheduled to commence on July 2,1977.

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The licensee subsequently performed an approximate best estimate analysis of the postulated worst case small pipe break, which included assumptions based on actual facility equipment availability during Core XII operation. The results of this analysis indicated that the calculated peak fuel cladding temperature was well below 10 CFR 50.46

limits, The.more conservative 10 CFR 50 Appendix K reanalysis of Core XII operation, however, indicated that 10 CFR 50.46 limits rnight have been exceeded in the ~ event that the safety injection pipe break had actually occurred.

Prior to returning the plant to operation after refueling of Core XIII the licensee: 1) performed flow measurements tests to determine the actual flow resistance through the safety injection piping; 2) changed the flow resistance in the safety injection' lines, by an ECCS modification; and 3) analyzed appropriate pipe break accidents in accordance with 10 CFR 50 Appendix K criteria.

The chanoes and results of tests and analysis were submitted to the NRC and were approved prior to restart of the plant after the refueling.6-7 DIESEL GENERATOR TRIP

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During a loss-of-pcwer test on August 26,1977, the E-4 diesel of

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the Peach Bottom Atemic Power Station Unit 2 started properly as a result of the undervoltage condition, but tripped immediately.

This trip was caused by the overspeed mechanism. The circuitry was reset, an adjustment was made to the mechanical governor to limit the diesel speed during a start and the unit was started successfully.

Because the exact cause of the trip was not firmly established, surveillance testing of the diesel was increased from once a week to once per shift.

During one of these tests, on August 27, 1977, the diesel tripped again.

Another adjustment was made to the mechanical governor, the load capability was checked and several successful starts were perfomed.

Once per shift surveillance was continued.

On August 29, 1977, the diesel again tripped on overspeed and was declared inoperable.

The diesel was then operated in excess of synchronous speed in order to determine the exact speed at which the overspeed mechanism would function.

This test determined that the diesel would trip at 940 rpm instead of the desired setpoint of 990 rpm.

The trip mechanism was adjusted to 985 rpm by a manufacturer's representative and diesel was started twice, successfully.

Investigation into the caus. of the change in the trip setting e

determined that during the diesel maintenance in June 1977 a camshaft

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In order to replace this camshaft the overspeed mechanism

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had been removed. When the overspeed mechanism was replaced, some

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necessary shims were not installed. Although this was the only diesel requiring this maintenance during the annual check, the other diesels were operated up to' a speed of 945 rpm to verify proper operation.

1:3 None of these diesels tripped on overspeed.

Analysis of this event revealed that a deficiency exists in the

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maintenance procedure associated with the diesel yearly inspection and the post-maintenance testing procedure. These procedures will be revised to correct the deficiencies.8 ELECTRICAL FAULT On July 13, 1977 while the personnel at James A. Fitzpatrick nuclear power plant were conducting refueling operations a short in a cable caused 600 volts AC to be introduced into a 115 volt circuit.

The 600 volt AC supply for the refueling bridge and the 115 volt AC circuit for refueling interlocks are both located in the same cable.

Flexing of the cable with bridge motion over the core caused the cable tc short internally.

The introduction of the 600 volts into the 115 se volt circuit caused nineteen relays in the rod manual control system te burn out.

All of the refueling operations were halted until the interlocks were repaired.

The rod worth minimizer and rod sequence control systems 'were also checked for damage.

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,7 A mcdification is being prepared that will remove the 115 volt AC~

interlo:k circuit from the cable carrying the 600 volt AC supply.

This will prevent recurrence.9 PIPE CRACK The Brur.swick Steam Electric plant Unit 2 was in hot shutdown and preparations were underway to startup the unit when the Shift Foreman noticed a snall leak of the recirculation loop suction piping.

This discovery was made during the closecut inspection of the drywell.

i Investication revealed.the leak was from a crack in the socket weld on a three-quarter inch test connection 900 elbow that was nonisolable, and the plant was placed in the cold shutdown conditicn.

The cracked pipe was cut out of the system and the connection was capped.

Similar connections on both Units 1 and 2 were dye-penetrant checked with no other indications of c'acks.

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Further investigation revealed that the crack was contained in the weld metal and intergranular stress corrosion in the heat affected zone of the base metal was ruled out.

A dye-penetrant inspectio'n of the internal and external diameters of this section of pipe revealed 4

no other cracks.

The inspection of the internal diameter of the socket weld joints shewed that a proper gap was present between the socket and the pipe end.

Based on a stress analysis and th'e observed condition of permanent defonnation of the failed area, along with the location of the crack, it is concluded that the initial crack was caused by stress concentration in the weld fillet area.

It is believed that this deformation was the result of worknen (during construction) using the pipe as a step.

This use of the pipe for this purpose plus vibrational stress resulted in the failure.

l A visual inspection of similar piping on the other loop of Unit 2 and both loops of Unit I revealed no deformation as was observed on the failed pipe.

It was also noted that the location of the three remaining pipes is such that they are not likely to be used as a step or support because of physical interferences.

These three pipes will be supported to protect them from experiencing excess: ge)yxternal yp loading and vibration, or will be removed and capped, t

a Point of

Contact:

l Joseph I. M Millen i

Office of Management Information and Program Centrol U.S. Nuclear P,egulatory Commission i

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REFERENCES E

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LER 77-2, Docket No. 50-285, January 31, 1977.

2.

LER 77-16, Docket No. 50-346, October 7,1977.

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Supplement to LER 77-16,' Docket No. 50-346, November 14, 1977.

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LER 77-23, Docket No. 50-344, July 29,1977.

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LER 77-21, Docket No. 50-219, September 23, 1977.

6.

LER 77-30, Docket No. 50-29, August 3, 1977.

7.

Sumary of June 17 Meeting, NRC-YAEC, June 22, 1977.

8.

LER 77-37A, Docket No. 50-277, September 9,1977.

9.

LER 77-43, Docket No. 50-333, August 11, 1977.

10.

LER 77-7, Docket No. 50-324, February 28, 1977.

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Suppleraent to LER 77-7, Docket No. 50-324, September 30, 1977.

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UNITED 3TATES NUCLEAR REGULATORY COMMISSION

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Those on Attached List COMPUTER LISTitlGS OF LICEt:SEE EVEtiT REPORTS SORTED BY FACILITY The enclosed computer listing provides information concerning licensee event reports entered into the file during the month of itovember.

If you desire additional information or special searches, please do not hesitate to contact us.

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f.'I.brk,ActingDirector Regulatory Info. Systems OfvTsion Office of t/anagement Infermation and Program Control

Enclosure:

As stated DUPLICATE DOCUMENT Entire document previously entered into system under:

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No. of pages:

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e Ol'C 09, 19ff tER MONiill Y Ottirtli SnRTED nY F ACIL I TY PAGE IB PROCE 5SEn DURING NOVEMnER FOR POWER REACIORS 1

FACllllV/SYSIEM/COMPONENf/

DOCKFT NO./

CVENT DATf/

COMPOHCHI sisnCODE/CAUSE CnDE/ t rR No./

nrenRT DATF/

EvCNT DESCRIPTION /

CAUSC SunLOOL/NANUPACIORLR CONIMUL NO.

MEPUMI fYrl CAUSL UCSCRIPTION DAVIS-nESSE-!

0's000346 090177 (N?-33-77-F68 Af 1580 HR$ UN 9/l/77 REACTOR COOLANT PUMP 1-2-1 TRIPPE0 000! ANT RECIRC SYS

  • CONTROLS 77 03L 80047T DUE in L OW SEAL filJECTION C COMPONENT COOLING WATER FLOW.

IN ACTION STA PUNPS 019176 10-DAY TEMEN! OF T.S.

3.4.1. WHICII REQtflRES 4 RCPS IN N00E REACTOR COOLANT

' Nll SunCOMPONENT PROVIDED PUMP l-2-1 RC-STARTLD AT 1524 HOURS Dr8 9/l/T7.

PTRSONNEL [RROR C Atl5 E SU0CODF NOT P fl0V 10 E 0 ITEM NOI APPL. I C AnL E PFRSONNEt EllRllit. AN OPER AIOR HISTAKENLY CLOSE3 VALVE I A tot WillCII SUPPL SE% INSTRUMENT AIR 10 NORMAL MAKEUP TLOW CnNIRnl VALVE.

DAVIS-nESSE-l 05000346 09077T (NP-11-77-741 Ai 0010 HR$ ON 9/7/77, SWITCHING FROM NORMAt TO ALTERNATE COOL A N T RECIRC SYS +. CONTROL $

77 0 3L 800477 POWER SUPPLY FOR DECAY DEAT SUCT I ON V ALVE DH12. V ALVE CL 0iED. REACTOR 0 VALVES O!9447 10-0AY PFRAIUM IPMFDI ATELY RE-OPENED VALVE C REESTAnLISHED Flow A00VE 2000 GPM.

NO SU0 COMPONENT PROVIDE 0 DESIGN / FABRICATION FRROR CAUSL SURCODE NUI PitOVIDEU liEM NOT APPLICA0LE OFSIGN DEFICIENCY IN INTERLOCK / CONTROL CIRCtlli CAUSED Till5 OCCURRENCE.

DAVIS-nESSE-l 05000346 09tFTT I NP-3 3-17-I s d AT 10001845 04 9/17/77 ' MAKEUP PUMP l-2 WAS REMOVED FROM REAC COOL CLEANUP SYS

  • CONT 11 O lt 1012FF SERVICE TO ALLOW MAINTENANCE TO REPAIR AN Olt LEAK ON 00100Aa0 DEARING E PUMPS 089374 30-DAY ND PLAIE.

IN ACTION STATEMENT OF TECil SPEC 3.1.2.4 SINCE 2 MAKE UP PUM NO SunCOMPONENT PROVIDED PS REQUIRED IN N000 3.

DE SIGN /F A nR IC Al l ON [RROR CAUSE SunC00E NOT PROVIDED IINGilAM PUMP CO DESIGN DEFICIENCY.

BUNA-N 0-RING ON REARING FND PLATE DIDN'T FORM ADEQU ATE SEAL AT DCARING It0USING.

DAVIS-DESSE-l 0$000346 092477 IIALF TRIP OF STE AM C FEEDWATER RUPTURE CONTROL SYSTEM CAUSED RISE IN REA O l l'R INSI SYS RE00 FOR SAFETY 77-016/Olf-0 100777 CIOR COOLANT SYSTEM TEMP c PRESSURE.

CAUSED PRESSURIZER POWER RELIEF VA I

INSTRUMENTATION

  • CONTROLS 019300 2-WEEK LVC 10 OPEN C VALVE FAILED TO CLOSE, CAUSING REDUCTION IN RCS PRESSURE.

U Tite R LC05 WERF EXCEEDED FOR 5 T.S.,

3.4.1, 3.4.5, 3.4.6.2, 3.6.l.4 C 3.7.1.2

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NOT APPL IC AIIL E CONSOLIDATCO CONTROLS CORP.

HALF TRir CONDlflDN FROM SFRCS CHANNEL 2. WillCH CAUSED VALVE FWSP7A TO C Cxid,C AUSE OF Tills IIALF TRIP llAS NOT DEEN POSITIVELY DETERMINED ALTil00 LOSE.

JSIVE INVESTICATION HAS REVEALED LOOSE CONNECTIONS AT TERMINAL 00 Gil ARDS IPOS $1nLE CAUSEl.

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cLEN ELLYN. ILUNOIS 60137

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NOV 221977

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Docket No. 50-346 Toledo Edison Company AT3: Mz. James S. Grant Vice President - Energy Supply Edison Plaza 300 Madison Avenue Toledo, OH 43652 Gestleaen:

This refers to the inspection conducted by Messrs. T. N. Tambling and T. L. Harpster of this of fice on September 26-30; October 5-7, 18-21, and 27,1977, of activities at Davis-Besse Nuclear Power S ta t io:,, Unit 1, authorined by NRC Operating License No. NPF-3 and to the discussion of our findings with Mr. T. Murray and mechers of your staff at the conclusion of the inspection.

The enclosed copy of our inspection report identifies areas exanisede during the inspection. Within these areas, the inspection consisted of a selective exatination of procedures and representative re cords, obs erva tions, and interviews with personnel.

During this inspection, certain of your activities appeared to be in noncocpliance with NRC requirements, as described in the enclosed Appendix A.

This notice is sent to you pursuant to the provisions of Section 2.201 of the N2C's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations.

Section 2.201 requires you to submit to this office within tventy days of your receipt of this notice a vritten statement or explanation in reply, including for each item of non-conpliance: (1) corrective action taken and the results achieved; (2) correcti@e action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved.

As discussed during the exit interview, it is requested that you submit a final follovup report on the September 24, 1977 event.

This report should include the chronology of events, pertinment transient data, evaluation of the transients and any long term effec ts, results of any testing and short and long term corrective action.

This report vill serve as a basis for a generic review of unusual transients.

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In accordance with Section 2.790 of the NRC's " Rules of Practice,"

j Fart 2, Title 10,. Code of Federal Regulations, a copy of this letter,

. the enclosures, and your response to this letter vill be placed in the b3C's Public Document Room, except as follows.

If the enclo-sures contain inforr.ation that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty days of your receipt of this letter, to withhold such inforr.ation fro:s public disclosure.

The application must include a full statement of the reasons for which the information is considered proprietary, and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application.

We vill gladly discuss any questions you have concerning this inspectio:.

S ince re ly,

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j Caston Fiorelli, Chief i

Reactor Operations and I

Nuclear Support Eranch Enclosure s :

1.

Appendix J., Fotice i

of Yiolation 2.

IT Inspection Report No. 50-346/77-32 cc w/ encl:

Central Files Reproduction Unit NRC 20b i

PDE local PDR NSIC

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_l Appendix A NOTICE OF VIOLATION Toledo Edison Company

  • Docket No. 50-346 E-ased on the inspection conducted September 26-30, October 5-7, 18-21, and 27.1977, it appears that certain of your activities vere in nonccepliance with NRC requirements below. The ite::t is a deficiency.

Contrary to the approved Quality Assurance Manual and Criterion V of 10 CFR 50, Appendix B, Administrative Procedure 1823.00 vas not cocpletely adhered to in the logging and review of jumper - lif t vires.

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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III Report No. 50-346/77-32 ~

Docket Ko. 50-346 License No. NPF-3 Licens ee:

Toledo Edison Company Edison Plaza

'300 Madison Avenue Toledo, OH 43652 Facility nace: Davis-Besse Nuclear Power Station, Unit 1 Inspection at: Davis-3 esse Site, Oak Harbor, OH Inspection conducted:

Septenber 26-30, October 5-7, 18-21, and 27, 1977

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Inspec tors :

T. N. Tambling

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T. L. Harpste:'

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Other Accompanying Personnel:

k'. Little, September 30, 1977 L. Engle, September 30, 1977 V. Leung, September 30, 1977 A. Szuklevicz, September 30, i' i7-J. Mazetis, September 30, 1977 J. Rajan, September 30, 1977 J. Pittman, October 19-20, 1977 A. Plumber, October 19-20, 1977 R. Denning, October 19, 1977

'II I/77 Approved by:

R. C.

p Ch Reactor Projects Section 1 E @ @.L.I 7 6 3 4 L b

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Inspection Sn m ry Inspection on September 26-30, October 5-7, 18-21, and 27, 1977 (Report No. 50-346/7 7-32)

Areas Inspected:

Investigated the causes, evaluation and corrective action associated with the sudden depressurization of the reactor coolant system on Septe=ber 24, 1977, routine, unannounced inspection of plant operation, tour of plant areas, follovup of modification to electrical grid system to meet licensee commitment, and nonroutine event reports. The inspection involved 110 inspector-hours onsite by two NRC inspectors.

Results :

Of the four areas inspected, no items of nonconplia'nce or devi-a tions were found in three areas; one apparent item of noncocpliance and two unresolved itees were found in one area (deficiency - failure to properly icplecent procedure for jumper-lif ted wire - Paragraph 6.b, and un:esolved - open water tight door, - Paragraph 7 - apparent defect in a cable penetration seal, Paragraph 7).

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DETAILS 1.

Persons Contacted J. Evans, Station Superintendent

  • I. Marray, Assistant Station Superintendent
  • L. Stalter, Technical Engineer
  • W.

Green, Adninistrative Coordinator

  • J. Ecck, Operations QA Manager
  • W. Derivan, Acting Operations Engineer L. Crime, Reliability Engineer D. Eriden, Chemistry and Health Physicist F. Faist, BW Site Operations Manager
  • Denotes those attending the exit interview.

Tae inspector also talked with and interviewed other licensee enployees including members of the technical and engineering staffs, reactor shif t crevs, and startup test leaders.

The inspector also participated in a meeting on September 30, 1977 at Davis-3 esse that included representatives of NRR, TECo Engi-neering, TECo Corporate Managenent, Babcox-Wilcox Company and Eachtel Corporation.

2.

Licensee Actibn on Previous Inspection Findings (Closed) Noncompliance (50-346/77-16):

Failure to properly doc.u-cent a review of a reportable occurrence and proper adherence to the administrative procedures for processing deviation reports.

The inspector found that Adninis trative Procedure AD 1807.00 was revised to clarify the apparent avkvardness in the procedure and that the procedure is being implemented to insure required review of reportable and nonreportable occurrences tracked by the devi-ation report system.

(Paragraph 6.a) 3.

Loss of Steam Generator,Feedvater Supply and Depressurization of Keactor Coolant Syste #

On Septecher 25, 1977, the licensee reported to Region III that a spurious trip signal in the Steam Feedvater Rupture Control System (SFRCS) on September 24, 1977 initiated a series of events that resulted in the loss of feedvater supply to the No. 2 Stean Gener-ator, de, pressurization of the Reactor Coolant System (RCS) and

_1] Licensee submitted 14 day Licensee Event Report NP-32-77-16 on October 7, 1977.

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.1 rupture of the rupture disc on the pressurizer quench tank that resulted in damage to the mirror insulation on the No. 2 steam gene ra tor.

An NRC Region III inspector was dispatched to the site September 26, 1977 to investigate the results of the incident, the action being taken by the licensee and corrective action planned.

Eased upon the inspectors review and upon telephone conversations between Region III and representatives of Toledo Edison Company, an imsediate action letter was issued to the licensee on Sep tenber 30, 1977.

This letter designated the corrective action required before the reactor could be returned to operation.

1 1

In the initial review of the incident, the inspector reviewed current status of the plant, proposed corrective action, details of the event, its safety significance, operation of engineered safety features during the event, conformance of the liniting conditions of operaticus, possible generic aspects, and possible radioactive releases or contacinations.

In followup, the inspectors reviewed the licensee evaluations of the incident, the results of testing and completed and/or planned corrective action.

The findicgs are as follows:

a.

Transient Chronology ~

Initial Conditions The reactor was at the 15% plateau in the startup test program and had accuculated approximately one effective full power day (EFPD) his tory. A high pressure turbine pressure tap between the turbine and the governor valve was found to be cracked. Power was reduced to approximately 9: and the turbine vas shutdown to repair the leak.

Main feed pucp turbine 1-2 was receiving steam from steam generator (OTSG)

No. 2, and feeding both OISG's through their respective startup feed control valves.

Secuence of Events 21:34:20 - A spurious half trip actuated the Steam-Feedwater Rupt.ure Control System (SFRCS).

The half trip closed the startup feed control valve to OTSG No. 2.

The operator has 8

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fr-only valve position demand signal indication and thus was unzware of the feed isolation. Other valves that would move on the half trip were already in the tripped position and thus gave no alarm.

The alarms received by the operator vere main steam lines 1 and 2 isolation valve (MSIV) solenoid trouble alarms which alars only on the conputer.

(This was a partial arcing of the control circuit that actuates the MSIV to close 'on a full STRCS trip).

21:34:44 - Low level alaru OTSG No. 2.

(Setpoint 24" startup range).

EFP turbine 1-2 is still steaming off the generator bu t the feed is isolated causing the level to decrease rapidly.

-21:34:56 - High temperature al' arm loop 2.

(Setpoint 560.6 F

. vide range cold leg). Decreasing OTSG No. 2 level reduces beat transfer capability.

21:35:16 - Eigh pressurizer level alarm.

(Se: point 220").

Coo lant is expanding into pressurizer from increasing loop tecperature.

21:35:18 - OTSC No. 2 lov level trip alarm.

(Setpoint 17" startup range).

This sigcal combined with the spurious half trip completes the logic for a full SFRCS trip.

The full trip closed the M31V's and lined up both OTSG's to the auxil-

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.tary feed systen.

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21:35:26-:49 - A reconstruction of data indicates that the.

pressurizar power relief valve actuatcJ and cycled nine times be fore f ailing in the open position.

The relief valve cycled about the setpoint (2255 psig) because a close relay which provides a 50 psi deadband was physically missing from the sy s t es.

The rapid cycling apparently caused some deformation of the pilot valve stem. However, pilot valve failed open due to galling of the ste:.

This resulted in the electro-catic relief valve failing open and led to a continued depres-surizatioc of the reactor coolant system (RCS).

21:35:36-:38 - AT?T's 1 and 2 discharge valves were open.

AFRT 2 only came up to 2600 RPM (normal is 3600 RPM) because of binding in the woodward governor linkage.

This corresponds to a shutoff head of approximately 700 psid.

Thus, no water vas fed into OTSG No. 2 as this pressure was considerably belov OTSO No. 2 steas pressure until 11 to 15 minutes into the transient.

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21:35:55 - Pressurizer power relief temperature high.

(Set-point 200 F).

Control room indication that relief valve had opened.

21:36:04 - AFP's 1 and 2 discharge valves open and water is being fed into OTSG No. 1.

21:36:07 - The operator manually tripped the reactor because pressurizer level was approximately 300" and rising.

About two seconds af ter the trip, level reached 303" and started de creas ing. Loop 2 hot leg temperature reached a maximum of approximately 584 F six seconds af ter the trip and started

. de c rea s ing. Loop 2 cold leg temperature reached a maximum of approximately 579 F 14 seconds af ter the trip. RCS pressure continued to blov down and various Reactor Protection Sys t em (RPS) trips occurred from low pressure as designed.

21:37:17 - The Safety Features Actuation System (SFAS),

incident level 1 initiated at 1600 psi.

The pressurizer quen:b tank vent isolation valves closed on containment isolation due to SFAS actuation.

21:37:49 - At this time it wa's noted that HPI flow indicator FYIH73A vas not indicating flow into the RCS, however, it was later determined that the initial flow was blocked by #~

two higher head cakeup pucps injecting 140 GPM through this line.

21:40:22 - The containment normal sump pump came on indi-cating the quench tank rupture disc had blown. HPI pumps vere shutdown at this time as pressurizer level was normal.

21:41:50 - Saturation pressure was reached in the reactor coolant sy s t em.

Steam formation was probably occurring in the Reactor Coolant Pucp (RCP) suctions.

21:43:41 - RCP's 1-1 and 2-2 were tripped.

At this time the transient was essentially terminated with the exception of,,the subsequent recovery actions.

The block valve for the failed electromatic pressurizer relief valve was closed approximately 20 ninutes af ter the start of the incident.

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Analysis of the Failures and Corrective Action (1) _SFRCS Half Trio The half trip logic in the SFRCS is used on certain valves in the steam and feedvater systems to meet the single failure criteria for isolation of the atmos-pheric steas lice power vent, the MSIV bypass, main steam varmup drain and startup feed control valves.

A one of four input vill close these valves.

(It should be noted that all except the main steam atmos-pheric power vent valve are normally in the closed position when operating above 15: pover).

A two of four logic is used on other valves in the Steam-Feed-wa t e r Sy s t e=.

Failure to see this spurious trip that initiated the incident was due to two reasons.

The logic circuit requires caly a signal duration of 25-35 milliseconds to lock in.

The cor.puter which is used to show the alarm condition is s:anning at a one second interval.

Th e re fore, trip signals of less 'than one second may not be seen on the computer. Other visual or alarm indi-cations (such as annunicators) were not.available in n' the control roce.

Corrective action by the licensee was divided into two phases to rectify this problem.

The first and immed-iate action was to connect six channel Brush Recorders on SFRCS input channels to provide detection of short cere spurious signals.

Follevup action involves the installation of annunicator vindows for half trips on s team generator lov level, loss of reactor coolant pucps I

and steam-feedvater delta pressure.

An annunicator j

presently exis ts on low steamline pressure.

A feasi-bility study will be made to determine whether a time delay circuit can be used on the trip signal to make

, sure that the computer vill also see short term signals (i.e. less than one second)

(2) nuxiliary Feedvater Pump 1

On a full SFRCS trip the two auxiliary feedvater pumps (AFP) are aligned acd started to feed the two steam.

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generators. The alignment is one AFP to one steam generator except for the main steamline t*eak accident.

For the stea:line break (detected by lov steamline pressure) the AF? to the af fected steam generator is aligned to receive steam and feed the unaffected steam generator (the one without the line break).

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When the spurious half trip isolated feedwater flow to No. 2 steam generator, a full SFRCS trip was initiated on lov steam generator level (see Transient Chronology)..

The AF?'s were properly aligned and started..However, No. 2 AF? only reached approximately 2600 rpo (vs desired 3600 rpc) due to a binding in the Woodward governor on the Terry Turbine driving the AFP.

At 2600 rpo the

=aximum pu=p discharge pressure is approximately 700 psig.

This head was not adequate to provide feedvater to the No. 2 steam generator until pressure in the steam gener-ator decreased to this pressure (approximately 11 to 15 minutes into the evect).

The loss of one st'ea: generator for a controlled cooldown of the reactor is within the accident analysis which as sc=es only one s tet= generator available.

How eve r,

the failure of the governor on the AFP turbine presents a generic, or cce=on code failure problem.

(This was je reported by the licensee in a letter from L. E. Roe to J. G. Teppler dated C:tober 11, 1977 in accordance with 10 CFR, ? art 21.21(b)).

The licensee through the manuf acturer analyzed the failure mechanis: of the governor.

It was concluded that under certain conditions the servocotor control driving the turbine speed control against the high speed stop places a misalignment force on the T-bar of the governor linkage.

This cisalig cent force creates a potential for the governor to bind at a speed position less than design s' peed upon turbine startup.

The governors for both AFP's

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were codified to correct this problem.

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~*- The failure of the No. 2 AFP to come up to speed and feed the No. 2 steam generator also resulted in the steaa generator " boiling dry."

This was concluded based upon the rapid rate of pressure decay inside the

- steas generator. Although this condition is not desir-able, the ic:ident is within the design analysis for the

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stean generator as supplied by the manufacturer.

The analysis placed a 20 cycle limitation on the generator over its lifetice.

The licensee has an ad=inistrative syste= for tracking operational transients over the lifetime of the plant to insure cycle limitations are not exceeded.

(3) Reactor Coolant System Depressurization The transient on the secondary side caused a corres-ponding operational transient in the Reactor Coolant Sy s tem (2CS).

This transient, while not desirable, t

vould normally be within the design capability of the syte: without danage to equipment.

However, due to the failure of the electromatic relief valve on the pressuri:er, to properly reset after relieving there was da= age to the mirror insulations on the No. 2 sters geserator, mi'..or damage to a ventilating duct and spillage of reactor coolant inside the containment vessel.

The licensee's investigation into the failure of the electrocatic relief valve revealed that the close relay was =issing from its control circuit.

The missing relay c.aused the relief valve to cycle around its sete' poin: of 2253 psig until the pilot valve steam stuck in the open position.

This failure of the pilot valve caused the relief valve to remain open continuously relieving the pressurizer to.he Quench Tank. The relief valve remained until aporoximately 20 ninutes into the event when the operatcr closed the block valve

to t':.e relief valve. Lack of earlier recognition that the relief valve had failed open was due to the fact th a t the operator did not have positise indications of the valve position on the control boacd. To correct this problem, the licensee installed a position light on the control board to indicate the position of the pilot valve solenoid.

(Usual indications that the relief valve opens initially is by a temperature monitor in the valve line).

The licensee's inspection of the pilot valve revealed that the sten stuck in the open position due to

" galling.

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There is no explanation as to why the reset relay was missing from the control circuit.

A review of die pre-operational test procedure performed prior to and during hot functional testing shows that the electromatic valve functioned properly.

Prior to the preoperational test, the licensee performed a yellov line (circuit checkout) of the control scheme. Both of these items indicate that the control relay had been in the circuit. It should be noted that the electromatic relief valve control circuits are not classified safety related and therefore do not fall within the normal quality control purview.

After inspection of the relief valve and replacement of the pilot valve, the 1.icensee tested the valve by manually cycling the valve six times with the RCS at approxica tely 600 psig. The pilet valve stuck in the open position again af ter the sixth cycle.

Inspection of the pilot valve stem revealed soce scratches on the ste= and that the outside diameter of the stes was.0005 inches oversize.

(Normal annulus clearance is

.001 inches). At the recommendation of the valve manu-facturer, the stroke of the solenoid for the pilot valve was re'duced from 3/8 inches to 1/8 inches.

This changefr in stroke still allows the pilot to function as designed and considerably reduces the stem surf ace area exposed to the steam, dirt, boric acid, etc. when the valve operates.

Reduction in surface area exposed to the steam flow prevents possible accuculation of contaminates on the beario; surface of the stem.

After the correction to the stroke travel and stem diameter, the electromatic relief valve was retested.

The valve was canually cycled ten times at approxi-nately 600 psig RCS pressure and once at approxima tely 2200 psig. The valve functioned as designed.

Ile RCS co=ponents are designed for forty cycles of a generalized depressurication transient in which the pressure drops 1400 psi and the temperature drops 61 F in fif teen minutes.

In the actual transient on the RCS side, the pressure dropped approximately 9

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s 1300 psi in eight minutes and the cold and hot leg tenp.

erature in loop 2 dropped approximately 41.5 and 45 F respectively in 7.5 minutes.

An analysis of the transient was performed by B&W for the licensee. Based upon this analysis, B&W concluded that this transient was within the scope of the gener-alized depressurization transient previously analyzed.

(4) Reactor Coolan_t Pu=ps Because the Reactor Coolant pumps (RCP) operated at or near the saturation pressure during portions of the transient, there was some concern of possible damage to puty shaf t seal bearings and impellers.

The operating condition was reviewed by the pump manufacturer.

The manufacturer concluded that there was small risk of any d a=a ge.

To provide assurance, the licensee performed instrumented tests in Mode 5 and 3 to verify normal operating paraceters.

(5) Reactor Fuel B&W also evaluated the possible ef fects on fuel perform -

ance as a result of the transient and concluded that there were no safety concerns with respect to the reactaf fuel. This conclusion was based upon:

Core burnup on September 24 was approximately 1 EFPD (no significant fission product inventory).

Because of the lov operating power history (15% and less) there was no significant decay heat source as compared to the source from the RCp's.

Conservative esticate that the maximum AP between the internal fuel rod pressure and RCS pressure was 300 psi and a maxinum clad temperature of 550 F a: this a?.

4 (6) ~ Containment Contamination from RCS The spillage of RCS vater did not constitute an airborne release problem inside containment.

This was due in part 4,

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because of the short operating history of the fuel assemblies. No detectable fission gases were present in the RCS.

There was low level contaminates due to normal activation products in the RCS (activation of norcal corrosion products).

This contamination was con-trolled and cleaned up by the licensee using standard radiation control procedures.

(7)

Training and Retraining To insure that operating personnel u'nderstood the se-quence of events, the licensee conducted retraining on the SFRCS. This training involved two four hour ses s ions. The first session involved a description and analysis of the event.- The second session covered a detailed description of SFRCS.

The training of the various shif t teacs plus other personnel was completed October 22, 1977.

4.

Second Loss o,f,Feedvater Transient On October 22, 1977, a spurious half trip from SFRCS closed the startup feedwater valve to the steam generator.

During this t rans i ent, all plant operating equipment performed as designed.

Both AI?'s started and reached 3600 rpm.

The pressurizer elec--

trosatic relief valve actuated twice and reset as designed.

Although the 6 channel 3 rush recorders installed to record the source of the spurious signal did not record the event, the licensee was able by a process of elimination to isolate the possible sources.

Two buf fer acplifiers and three integrated circuit clips were replaced as probable causes for the spurious sigaal. The licensee is continuing his efforts to positively identify the source of spurious signal.

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5.

Elect rical Grid Stabil,ity Modifi,ca t_ ion Per Condition 2.C.(3)(q) of Operating License NPF-3, the licensee su' emitted an evaluation and proposed modifications to electrical grid protective system to NRR for review.

The purpose of these modifications is to insure adequate breaker coordination, alarm

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and isolation of the onsite electrical system in sufficient time to percit the required Class 1E equipment to operate in the event of offsite grid degradation.

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The inspector reviewed and examined the implementation of the subject modification as designated by the licensee's letter of July 18, 1977 (Seriti No. 293) to J. E. Stoltz from L. E. Roe

' and Facility Change Request 77-217 (original and supplement No. 1).

No deficiencies were identified by this review.

The review ef fort included review and examination of procurement records, certification for Class lE equipment, work orders used for installation, setpoint changes and setting, safety review, SR3 review of the facility change and procedures used, and dis-cussions with cembers of the engineering staff and operating staff involved in the design and installation of the modi.fi-ca tions.

6.

Plant Operations - General The inspector reviewed general plant operations including an examination of selected operating logs, jumper-lif t vire logs, deviation reports for the period of July 1977 through October, 1977.

This reviev was cade to determine compliance with tech-nical specifications and administrative procedure requirements.

1.

Deviation Reports While reviewing the deviation reports, the inspector noted that 38 reports (for the period July 11, 1977 to Augus t 16,1977)-

had not been filed in the caster file.

Seven of these were on the SRI agenda for final review and closeout.

The others were still outstanding for final resolution.

It was noted that although copies of these reports were not in the caster file, the reports are logged and are being tracked by the technical section. For the reports found in the master file, many still re=ain open and require final closecut.

In response to a previous item of noncompliance (Inspection Report 77-16), the licensee initially revised Administrative Procedure AD 1807.00 on July 12, 1977 and approved it for itplacentation on September 9,1977 to improve and clarify the review process for deviation reports.

It was noted there has been an apparent inprovement in the tracking and review of DVE's issued since 'the procedure revision.

Eevever, even though the DTR's are used by the licensee as a control docunent that does not get closed out until all 9

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corrective action is completed (including corrective action covered by facility change request or action item record),

there appears to be a large number of open D7R's in the files.

Many of these can and should be closed in a more orderly matter.

This need to reduce the large backlog of open D7R's was discussed with the licensee during an exit interview.

b.

Jeepers and_ Lift Wire Log Tne inspector reviewed the jumper and lif t wire log and the implecentation of Administrative Procedure AD 1823.00, Jenper and Lif t L' ire Control.

From a previous inspection, (Inspection Report 77-16), it had been noted that a large um=ber of jumper and lif t vire ' tags were outstanding.

The status of the licensee's effort to reevaluate the need of these jumpers and lif t vires was reviewed.

Considerable progress bad been cade and the effort is still in progress as noted by the internal review presently being conducted by the operations section.

The inspector selected several jumper-lif t vire - tags nesbers at rando= and verified that they do exist.

Tie need to expedite this effort was discussed in the ezit int e rview.

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r A3 1823. 00 jucper-lif t wire log sheets requires the persons placieg the tag to reference the work order request number a:d the reason for the tag on the log sheets.

A review of the leg sheets indicated that this was not being consistent 1 '

7 dene. Exanples were:

Tag No. 4073 through 4084, no reason was given for tags.

Tag No. 4163 through 4166, no work order number was re f e ren ced.

i Tag Ko. 4123 through 4126, no reason was given for tags.

In the exit interview the inspector discussed how failure to provide this information is considered an inadequate review by the tagging supervisor (DBTS), the Shift corecan who is suppose to review the current status each shif t and the Operations Engineer (or his representative) monthly reviev.

This failure to properly implement AD 1823.00 is considered an itec of noncompliance. -

k.

s Also discussed in the exit interview was a possible incon-sistency in how individual DBTS handled critical tags (tags with safety implicationt).

The current procedure does not require documentation to show how and what the D3TS considered in the placement of critical tags.

7.

Plant Tour Ibe inspector toured various areas of the plant to observe oper-ations and activities in progress.

This included general state of housekeeping, proper aligament of valves in the high pressure injection af SFAS, status of EVS boundary, leaks, pipe vibrations, radia, tion controls, shif t canning, discussion with operating per-sonnel concerning lighted annunciators, and review of a startup procedure currently being implenented.

No ite=s of noncompliance vere specifically identified. However, two items were lef t unresolved and there was one iten of major conc e rn.

As discussed in an exit interview, the plant is currently oper-ating with a large number of lighted annunciators.

The inspector stated that there appears to be a need for the licensee to review the current status of lighted annunciators for the purpose of eliminating nuisance alarms, alarms that apparently have logic g-problems or in which the setpoint span may be too tight.

Examples are:

Panel 1, vindov 2-8, Emerg D/G FOS T 1/2, Hi/Lov.

If tank is overfilled, light is lit (alarmed condition). The operators concern is lov level since he has no other means to detercine what the level is.

A sudden change in the status of tank could go undetected (from Hi alarm to Low ala re).

Panel 1, vindow 1-5 and 1-6, ESSEN Bus E-1 and F-1, Breaker not normal.

Vindev always lit because there is no norcal pos i tion.

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Pane 15, vindows 5-1, SFAS CTMT Rad Low '. ail. Windov remains always lit because of detector location.

There is not suf ficient background radiation to make the meter read above Zero.

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On October 27, 1977, at approxicately 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, the water tight door separating the two auxiliary feedwater pumps was found open during a tour.

The door was immediately closed.

This ite= is unresolved pending further investigation.

On October 27, 1977, a black oily material was found dripping from a ceiling cable. penetration in the cable spreading room.

This itec is unresolved pending further investigation by the lice nsee.

8.

Onsite Review of Inspection Process The inspector was accocpanied by tv5 people from Battelle-Columbus and one person fron the Of fice of Nuclear Regulatory Research on October 19 and 20,1977.

Battelle is currently under contract to the NRC to evaluate cethods for applying WASH-1400 methodologies to the inspection process. Observations were made of the current inspection process.

9.

?.eview of Fonroatice Ivents Reported by the Licensee The inspector revieeed licensee actions with respect to the folleving listed nenroutine events reports to verify that the events were revieved and evaluated by the licensee as required by Technical Specifications, that corrective action was taken i

by the licensee, acd that sa fety linits, limiting safety systen set t ings, tad limiting conditions for operation were not exceeded.

The inspector examined selected Operations Committee minutes, licensee irvestigation reports, logs, and records, and inspected equipment and intervieved selected personnel.

Tva inoperable relative position indicators in Group 6 of CRD Sys tem (SP-32-77-15).

Ioss of Reactor Coolant pressure due to f ailure of pressurizer power operated relisf valve (NP-32-77-16).

(See Paragraph 3 for details)

No it e=s of nonco=pliance or deviations were identified.

The folleving licensee event reports were reviewed and closed out on t'ce basis of an in-office review and evaluation:

Hydr $ test on Eigh Pressure Injection line (NP-33-77-28).

a.

b.

Steam generator level limit exceeded (NP-33-77-30)..

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Chlorine detector AI 5358A inoperable (NP-33-77-31).

c.

d.

Loss of Ehield 3uilding integrity for B Purge System 18 y

conth test (KP-33-77-33).

e.

Plan:ed replacement of overload heaters -in Control Room Emergency Vent condensing unit (NP-33-77-37).

f.

loss of DC peser to speed control switches for SS 815 and SS 816 for tes ting (KP-33-77-39).

g.

Ze= oral of Auxiliary Feedvater pump 1-2 from service to implement viring change (KP-33-7 7-41).

h.

Yain stea= supply line check valve bonnet leak (NP-33-77-42).

i.

Loss of shield building integrity due to unlatched door (KP-33-7 7-47).

j.

Coc t ain=es t nor=al sump total flow instrument string inoper-able (K?-33-77-49).

k.

AFP turbine 1-2 inoperable due to loss of speed control (K?-33-7 7-51).

1.

AFP turbi:e 1-1 inoperable due to loss o'f speed-control UE' (N?-33-7 7-52).

ATP 1-2 inoperable due to ground in speed control switch

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(N?-33-7 7-5 3).

n.

Flev pa th fro: bo:ic acid storage system operable (NP-33-77-54).

o.

Yain s t ea: isolation valve M 5100 failed closed due to loss of cc: trol air (N?-33-7 7-55).

p.

NI-3 inoperable to connect to reactimeter (NP-33-77-59).

q.

A?I for control rod 4 of group 4 inoperable (NP-33-77-63).

DH pu:p 1-2 made inoperable to install union in cooling water line (N?-33-77-66).

Chancal 3 of RPS inoperable (NP-33-77-67).

s.,.

e.

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p t.

Eydraulic snubber EED-19-E144 inoperable (NP-33-77-69).

ain steam line hydraulic snubber SR 17 and SR 11 inoperable u.

(N?-31-77-70),

1?P 1-1 stea= supply isolated for caintenance (KP-33-77-71).

v.

Eeactor coolant system T avg less than 525 F (NP-33-77-75).

v.

10.

U resolved Ites Utresolved ite=s are catters about which more information is required is order to ascertain whet.her they are acceptable items, itecs. of nonco:pliance, or deviations.

Two unresolved items dis-closed during the inspection are discussed in Paragraph 7.

11.

Erit Int erviev Tce inspectors pet with licensee representatives (denoted in Para-Eraph 1) o: October 7, 21 and briefly on October 27,1977 to su carize the findings of the inspection.

The licensee represen-ta:ives cafe the following remarks in response to certain of the ite=s disecssed by the inspector.

October 7,19 77

..f Stated that in addition to the corrective action taken to date on the Se;tember 24, 1977 incident they would (Paragraph 3):

I:strument SFRCS inputs to help detect spurious signals.

a.

b.

Add SFRCS acnunciator vindrvs.

S:udy the feasibility of a time delay mechanism so that the c.

c:=puter can log short ter: spurious signals.

d.

Crsplete training on the SFRCS by October 22, 1977.

e.

Tes t the codified AF? turbine governors in place in Mode 3.

f.

Test the electrocatic pressurizer relief valve cold and hot (at approritately 600 psig).

During a telecon on October 14, 1977, when the relief valve failed on the sixth hot cycle e -

L b

.: s tes t, the licensee stated that they would retest the valve ten times at 600 psig and once at approximately 2200 psig.

Complete the testing of the reactor coolant pumps at a g.

pressure equal to or above 1300 psig.

Acknowledge that the return to power operation was predicated upon the successfull cocp'letion of the above tests.

Also stated that they would keep the inspector informed of the progress of the tes ting.

Acknowledge the inspectors request for a detailed follovup report on the September 24, 1977 incident including a detailed analysis of the long term effect of the transients.

October 21, 1977 Acinowledge the inspectors concern about the jumper-lift wire logs and ackncvledged the inspectors statement with respect to the appa re:t ites of noncocpliance (Paragraph 6.b.).

Acinculedge the inspectors concern about the status of control roco a:aunciators and stated that they have been pursuing the proble:.

(?aragraph 7) f Ackncvledge the inspectors statements about the number of outstanding devia tioc reports.

(Paragraph 6.a. )

i Oc tob e r 27, 1977 Act:orledge the inspectors review of the modification of the elec-trical grid systec.

(Paragraph 5)

Ack:ovledge the inspecters findings concerning the open vater tight door and stated that they would check their procedures to determine why the door was lef t open.

(Paragraph 7)

Ack:ovledge the inspectors findisgs concerning the black oily caterial dripping froc a penetration in the cable spreading room and that they vould icmediately investigate and determine the extent of the problen.

Informed the inspector on October 28, via telephone that the problem appeared to be confined to the one pene-tration and that they were pursuing it further with both Dov Chenical Cocpany and BISCO.

(Paragraph 7)

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.3 Docket No. 50-346' # _ ' '

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JAMES S. GRANT v.c. am%

f aerer Swoon December 14, 1977

"'S' 858-2838 Serial No. 1-6 Mr. Gaston Fiorelli, Cqief Reactor Operations Branch, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137

Dear Mr. Fiorelli:

Toledo Edison acknowledges receipt of your November 22, 1977, letter and report enclosure 7-32 referencing an apparent deviation from Nvis-Besse Nuclear Power Station Unit No. 1 consitments to the NRC, listed as a " Deficiency" under the heading " Notice of Violations".

Following a thorough exanination of the item of concern, Toledo Edison herein offers information 'regarding this item, including corrective actions and the dates of corrective actions.

Item b Deficiency: Contrary to the approved Quality Assurance Manual ard Criterion V of 10CFR50 Appendix B, Administra-tive Procedure AD 1823.00 was not completely ad-hered to in the logging and review of jumper-lift vues.

Response

The corrective action taken and the results achieved ac:1 the corrective action taken to avoid further non-cc:pliance are as follows:

1.

Adainistrative Procedure AD 1823.00, Jumper and Lifted Wire Control, was revised November 10, 1977, to insure the jumper and lifted wire log sheets are filled out consistently.

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<J-Ss THE TCLEDO EDSON COMPANY ED: SON PLAZA 300 MADISON AVENUE' TOLEDO. OHIO 43652

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Docket No. 50-346 Serial No. 1-6 Dececher 14, 1977 2.

The Davis-Besse Tagging Supervisors (DES) have been inforned of the importance of filling out the jumper and lifted wire log sheets correctly and con,sistently.

Full compliance will be achieved when all of the outstanding jumper and lifted wire log sheets are updated, which will be accomplished by December.20, 1977.

1 Very truly yours,'

A I

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JSO/h'HG/ daw l

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