ML20045H744

LER 93-012-00:on 930615,automatic Reactor Scram & Isolation of Inboard Group 2 PCIS Valves Occurred.Caused by Loose Packing Nut on Instrument Isolation Valve.Valve Repaired & Similar Installations on Units checked.W/930709 Ltr

ML20045H744
Person / Time
Site:	Hatch
Issue date:	07/09/1993
From:	Beckham J, Beckman J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-3398, LER-93-012, LER-93-12, NUDOCS 9307210210
Download: ML20045H744 (10)

Text

. .' . Georgia Power Company 00 invernes Center Parkway -

Post Offse Box 1295 E -

Birmingham, Alabama 35201 p Telaphone 205 877 7279 m

J. T. Dockham. Jr.

Vice President - Nuclear Geoi$tt Potver Hatch Proioct the sout'>em ekt??c system July 9, 1993 Docket No. HL-306 50-321 005007 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant Licensee Event Report Instrument Isolation Valve Packing Leak Results in an Automatic Scram Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(iv), Georgia-Power Company is submitting the enclosed Licensee Event Report (LER) concerning an instrument isolation valve packing leak which resulted in an.

automatic reactor scram. This event occurred at Plant Hatch - Unit 1.

Should you have any questions in this regard, please contact this office.

Sincerely, N

. T. Beckham, Jr.

JKB/cr

Enclosure:

LER 50-321/1993-012 cc: Georaia-Power Company Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion Il Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert', Senior Resident Inspector - Hatch 9307210210 930709 7A fj PDR ADDCK 05000321 S PDR ,

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LICENSEE EVENT REPORT (LER)

FAllLily NAML (1) UULALI hbMbik (2) FW fan PIANT E. I. HATCH, UNIT 1 05000321 1 0F I 9 TilLE (4)

INSTR 1 RENT IS01ATION VALVE PACKING IEAK RESULTS IN AN AUlOMATIC REACIOR SCRAM EVEhI DATE (5) LER h0MBER (6) REFORT DATE (7) OlHER FACILillE5 INVOLVED (8)

MONTH DAY YEAR YEAR SEQ hum REV MONTH DAY VEAR FACILITV NAMES DOCALT huMBER(5) 05000 06 15 93 93 012 00 07 09 93 05000 OPERATING MODE (9) 1 20.402(b) 20.405(c) ^ 50.73(a)(2)(iv) 73.71(b)

POUER -

20.405(a)(1)(1) -

50.36(c)(1) -

50.73(a)(2)(v) -

73.71(c)

LEVEL 100 20.405(a)(1)(ti) 50.36(c)(2) 50.73(a)(2)(vii) OTHER (Specify in 20.405(a)(1)(iii) 50.73(a)(2)(1) 50.73(a)(2)(viii)(A) Abstract below) 20.405(a)(1)(iv) -

50.73(a)(2)(li) -

50.73(a)(2)(viii)(B) 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

LICEN5E E CONTACT FOR Th15 LER (li)

NAME TELEPHONE NUMBER 4REA CODE STEVEN B. TIPPS, MANAGER NUCIEAR SAFETY AND COHPLIANCE, HATCH 912 367-7851 COMPLETE ONE LINE FOR EACH F AILURE DESCRIBED IN THI5 REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC- P0RT CAUSE SYSTEM COMPONENT MANUF C- R PORT TUR ppg X JA ISV D232 YES SUFFLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR SUBMISSION DATE (15)

] YES(If yes, complete EXPECTED SUBMISSION DATE) ] NO AB5 TRACT (16)

On 6/15/93, at 1453 CDT, Unit 1 was in the Run mode at a power level of 2436 CMWT (100 percent of rated thermal power). At that time, an automatic reactor scram and isolation of the inboard Group 2 Primary Containment Isolation System (PCIS) valves occurred due to a false low reactor water level signal. Control rods fully inserted as designed. As expected, actual water level decreased immediately following the scram, reaching a minimum level of 34 inches below instrument zero (124.5 inches above the top of the active fuel). At approximately 10 inches above instrument zero, an actuation of the outboard PCIS occurred on an actual low water level condition. The Reactor Feedwater Pumps (RFPs) responded to the actual low level condition and restored water level. No Emergency Core Cooling Systems actuated as a result of the low water level condition, nor were they required to actuate. Reactor pressure decreased as a result of the scram and was then controlled by the Turbine Bypass Valves at approximately 920 psig. Due to misleading level indications, level increased above the bottom of the Main Steam Lines, resulting in water intrusion into the lines.

The cause of the event was a loose packing nut on an instrument isolation valve becoming disengaged during an instrument calibration. This ultimately resulted in depressurization of the sensing line and a false low reactor water level signal. Corrective actions include repairing the valve, checking other similar installations on both units, performing a walkdown of system piping, and analyzing the effect of water in the Main Steam Lines.

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" LICENSEE EVENT REPORT'(LER)

TEXT CONTINUATION FACILITY*NAME (1) DOCKET NUMBER (2) LER NUMBER (5) PAGE-(3)

YEAR SEQ hum REV PIANI E. I. HATQi, UNIT i 05000321 93 012 00 2 or 9 TEXT PIANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System '!

Codes are identified in the text as (EIIS Code XX).

DESCRIPTION OF EVENT On 6/15/93, at 1453 CDT, Unit 1 was in the Run mode at a power level of 2436 CMWT (100 percent of rated thermal power). At that time, an automatic reactor scram and automatic isolation of the inboard Group 2 Primary Containment Isolation System (PCIS, EIIS Code JM) valves occurred due to a false low reactor water level signal. >

L Calibration of reactor water level instrument 1B21-N093B was in progress at the time of the scram. 'This level transmitter providesLa trip signal to the Eigh .

- Pressure Coolant Injection System (HPCI, Ells Code BJ) on a high reactor' water

--level condition. The Technical Specifications require the calibration at Lleast once per 18 months. When the'nonlicensed Instrument.& Controls (I&C) technician I performing the calibration attempted to close a 3/8 inch' instrument sensing line isolation valve, the packing nut came off the valve bonnet upon contact with the  !

-stem handle. Subsequently, the packing gland and packing material partially came out of the bonnet resulting in a substantial bonnet leak which partially  :

depressurized the instrument line.

This instrument line is a variable, or low pressure-leg, serving various level transmitters that provide input to the Reactor Protection System (RPS, EIIS Code JE), the PCIS inboard valves, Reactor Water Level Indicators 1B21-R606A and C, i the "A" subsystem of the Feedwater level Control System (FWLC, EIIS Code JK),.as  !

well as other systems. Consequently, when the instrument line depressurized, a  ;

false low water level was sensed by these instruments, resulting in the reactor-scram, automatic isolation of the inboard PCIS valves, and a false low level i indication on Main Control Room indicators 1B21-R606A and C. The FWLC. System 'I was in ."B" control at the time and, . therefore,- was not affected by the false - '

signal.

The control rods fully inserted as designed. As expected, immediately following l

.the scram, reactor water level decreased due to void collapse in the. reactor- ,

coolant. The minimum water level reached during the transient was 34 inches below instrument zero (124.5 inches above thel top of the active fuel) before- -

level was' recovered by the Reactor Feedwater Pumps (RFP,.EIIS Code SJ). During' '

the level transient, at approximate 1y'10 inches'above instrument zero, the PCISL received a second automatic. isolation' signal on an actual' low' level condition, i resulting in closurelof the outboard PCIS valves.

It is apparent that following the initial.depressurization of the instrument i line, the packing partially sealed off the: bonnet leak. As.a consequence, the  :

line7 partially repressurized, resulting in the associated level instruments l'

. tracking reactor water level at lower than actual level. Specifically, Reactor Water Level Indicators 1B21-R606A and C indicated a lower than actual level, and  :

level instruments IC32-N004A and C, which input to the "A" subsystem of.FWLC ,;

and/or the Main Turbine and the RFP trip system, were also sensing a lower than '!

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TEXT CONTINUATION FACILITY NAME (1)' DOCKET NUMBER (2) LER NUMBER (5) PAGE (3)

VEAR SEQ hum REV PIMI E. I. HAIUI, UNIT 1 05000321 93 012 00 ~3 0F 9 IEXT actual level. The magnitude of this discrepancy between actual-and sensed level varied during the event.

During recon ry of reactor water level, RFP "B" was secured per procedure to-preclude overfilling the reactor vessel. Level continued to increase, and RFP-

"A" and the Main Turbine tripped on a high reactor water level condition. The actual level at the time of the trip was approximately 68 inches:above instrument zero. The trip setpoint is 54 inches. The delay in the trip system

-actuation was due to level instruments 1C32-N004A and C sensing reactor. water-level approximately 14 inches lower than actual due'to the partially depressurized sensing line.

When sensed water level decreased below the high level trip setpoint, RFP "A" was restarted and placed in automatic level control, aligned to the "B" subsystem of the W LC system. The "B" subsystem of WLC is' served by a separate instrument sensing line and, therefore, was not'affected by the failed instrument line. Consequently, WLC was sensing an actual high reactor water -

level condition, resulting in RFP "A" running on minimum flow and not injecting _

into the reactor vessel, i a

Reactor' pressure initially decreased to approximately 757 psig as a result _of ,

the scram and was then controlled by the Turbine Bypass Valves (TBV) at approximately 920 psig.

i Licensed operators, in responding to the event, monitored reactor water level. l They were aware of a problem associated with a "B" level instrument during -

y calibration that ultimately caused the scram. However, the affected instrument line was unknown and, therefore, the impact of the condition on their instrumentation was also unknown. Reactor Water Level Indicators 1B21-R606A and  ;

C appeared to be tracking level after the scram, and Reactor Water Level Indicator 1B21-R606B was'off scale high. The high end of the' scale is 60 inches above instrument'zero. -Licensed operators concluded from the displayed-level l indications that the "B" indicator had failed upscale as a result of the failed l sensing line and that the "A" and "C" indicators were accurately displaying l level.

The level instruments that provide input to the WLC System "B" control are served by the same sensing lines that provide input.to the "B" Reactor Water Level Indicator. Consequently,.the operators, questioning the accuracy of the "B" Reactor Vater Level instrumentation, transferred the WLC System from.."B"- y control to "A" control. The RFP was then being controlled by the "A" WLC subsystem and periodically injected. into. the reactor-vessel as a result of the sensed false low water level. Control room personnel believed at this point

.that reactor water level was being maintained.within an acceptable band by the RFP. 'In actuality, reactor water-level was high and continued to increase each time RFP "A" injected.

During this time, support personnel,. in conjunction with some of the shift personnel, were investigating the cause of the scram. At approximately 1520 CDT, they determined that the depressurized instrument sensing line served the i

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YEAR SEQ hum REV PIANI E. I.1%IQi, UNIT 1 05000321 93 012 00 4 0F 9 TEXT-

"A" and "C" Reactor Water Level Indicators ~and not the "B" indicator. -

Therefore, they. questioned the accuracy of the "A" and "C" indicators At that l point, the "A" and "C" indicators were displaying a level of approximately 28 ,

inches above instrument zero, but the actual level was approximately 121 inches' above instrument zero, as indicated by the floodup range instrument. .

-j Consequently, actions were taken to lower level. RFP "A" was secured,.and thef Reactor Water Cleanup System was aligned to the Main Condenser in order to ,

lower level.

The bottom of the reactor. vessel nozz1ss for the Main Steam Lines (MSLs). is at 111 inches above instrument zero. Consequently,.with the reactor water level  ;

greater than 111 inches above instrument zero, reactor coolant was entering the MSLs. It is believed that the coolant was then being diverted to the Main.  :;

Condenser via the TBVs and MSL drain valves. Procedure 34AB-C71-001-1S, " Scram  ;

Procedure," requires that, if reactor water level exceeds 100 inches above-instrument zero, the Main Steam Isolation Valves (MSIVs) should be closed. The' .i purpose for closing the MSIVs is to prevent damage to the lines downstream'of  !

the MSIVs and to the Main Turbine if water enters the MSta. During scram '

recovery, licensed management personnel made a conscious decision not to close ,

the MSIVs based on the following factors: 1) The Main Turbine had already tripped; therefore, water could -not enter the Main Turbine. 2) Closing the MSIVs would have complicated' scram recovery in that the normal reactor feedwater and the Main Condenser would be unavailable. 3) At the time the action was .;

considered, reactor water level had been accurately assessed and was decreasing.

At approximately 1550 CDT, the reactor water level had decreased below the bottom of the MSL nozzles, and water was no longer entering the MSLs.

The repair of the packing leak required isolation of the affected instrument ,

header which serves Emergency Core Cooling System (ECCS) instrumentation, as j

-well as the RPS and PCIS instramentation previously mentioned. Consequently, in "

accordance with the Technical Specifications, a Limiting Condition for Operation ,

requiring that Co?.d Shutdown be achieved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the condition.is not

-repaired was entered. By 0305 CDT, on 6/16/93, the condition had been repaired,

'the' instrument lins unisolated, and the Limiting Condition for Operation

-terminated. ,

CAUSE OF EVENT The cause of the' event was component failure in that ' a loose packing ' nut became j disengaged from an isstrument isolation valve bonnet during a maintenance activity. As described previously, when the nonlicensed Instrument & Controls: l (16C) technician performing a calibration on level instrument 1B21-N093B began: ,

, to close a 3/8 inch instrument sensing-line isolation' valve, upon contacting the

-stem handle, the packing nut came off of the bonnet. Subsequently, the packing.

. gland'and packing partially cameLout of the bonnet, resulting in a substantial '

bonnet _ leak. 'The associated instrument line serving various level transmitters i which provide input to RPS, PCIS, Reactor Water Level Indicators 1B21-R606A and  :

C, the "A" system of the FVLC, as well as other systems, depressurized. . .

Consequently, when the instrument line depressurized, a false low water level :l was' sensed by these instruments. resulting in a reactor scram, automatic ,

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IIANT E. 'I. HATCH, UNIT 1 05000321 93 00 5 0F 9 TEXI

. isolation of the inboard PCIS valves,'and a false low level indication on Main a

Control Room indicators 1B21-R606A and C. The FWLC System was in."B" control at the time of the scram and, therefore, was not initially affected by.the false signal. However, when control was transferred to "A" during scram' recovery, the .)

FWLC System controlled the RFP based on the false ' low water . level' signal, . +

ultimately resulting in the high reactor water level condition.

The cause of the high reactor water level condition was the partial

-repressurization of_the instrument sensing line. . Typically,z a sensing line

' failure would result in a total depressurization of the line withouc .

repressurization. -In such a situation, the instrument served by the line would ;l fail upscale or downscale and would not respond-to actual water level. changes, t Such was not the case in this event. It is apparent from a review of the Safety _.

Parameter Display System (SPDS, EIIS Code IQ) graphs that the packing leak j

. partially sealed off after the initial depressurization. The graphs show that.

-sensed level on the "A" and "C" instruments initially-went downscale. The  !

graphs show that level was restored and then nominally fluctuated as would be f expected due to.the coolant boil-off and periodic'feedwater: additions. This '

phenomenon, coupled with the "B"

~

indicator being upscale, led the- operators to l conclude that the "A" and "C" indicators were correct and the "B". indicator instrument referenceLline had depressurized causing it to fail upscale. An  ;

additional factor affecting.their conclusion was'that they knew a "B"_ level ,

instrument was being calibrated ~at.the time of the event.' The operators- j associated the "B" Reactor Water Level Indicator with the instrument being- . .

calibrated and surmised that a problem with the calibration had caused the "B" Reactor Water Level Indicator to fail upscale. .;

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REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT- I

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This report is required pursuant to 10 CFR 50.73(a)(2)(iv) in that it' involved .

j unplanned automatic actuations_of Engineered Safety Features (ESF). 1 Specifically, a false low reactor water level condition resulted in automatic-  ;

RPS and.PCIS actuations. Additionally, during the level transient following'the  :

scram, an actual low water level condition resulted in the outboard PCIS' valves .)

automatically closing.

The RPS provides timely protection against events that'could potentially result' in damage to the fuel by initiating an automatic scram when appropriate plant -l

-parameters exceed design limits. One of the plant conditions'that would result ]

fin lan automatic RPS actuation is a lowLreactor water level condition. A scram dd

'is initiated in this condition to reduce the heat generation rate of ' the ' fuel to ]

prevent fuel damage due to the reduced ~ coolant inventory and, thus, reduced v  !

cooling capacity. .!

'I In'this event, depressurization of an instrument sensing line resulted in two.

level-instruments, which provide input to RPS .failing-low and initiating a trip ];

in the RPS-logic. As designed, the two inputs were sufficient to trip the

. one-of-two-taken twice RPS logic. All control rods fully inserted as designed.

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TEAR SEQ hum REV PIANT E. I. HATCH, INIT 1 05000321 93 012 00 6 or 9 IEXT The PCIS provides automatic isolation capability of. Primary Containment penetrations to preclude the release of radioactive material and the loss of reactor coolant inventory in the unlikely event of an accident. The system is designed to actuate on a low reactor water level condition. The level instruments that innut to RPS also input to PCIS. Consequently, the false low l level condition raw 'ted in isolation of Group 2 PCIS valves. Only the inboard l valves closed due he false low level condition since only the inboard PCIS l is served by the lecal instruments on the affected instrument sensing line. The

! actual level decrease that followed the scram resulted in an actuation of the outboard PCIS. As a consequence, the outboard Group 2 PCIS valves received an L automatic closure signal. The PCIS valves were confirmed to have closed as L required.

Prior to the event, reactor water level was at the normal level of approximately 37 inches above instrument zero. As expected, immediately following the scram, l actual reactor water level decreased due to void collapse in the reactor I coolant. The RFPs responded to the actual decrease and restored level. The minimum level reached in the transient was 34 inches below instrument zero l (124.5 inches above the top of the active fuel). The initiation setpoint for j HPCI and the Reactor Core Isolation Cooling System (RCIC, EIIS Code BN) is 35 )

inches below instrument zero. Consequently, these systems were not required to  !

initiate and did not do so.

l I

Following the restoration of reactor water level, it continued to increase due

]

to RFP injection as discussed previously. The FWLC System is comprised of an "A" and a "B" reactor water level input, either of which can be selected as the reactor water level input to control the system. The "A" reactor water level ,

input is from level transmitter IC32-N004A, which is served by the sensing line J that depressurized in this event. The "B" level input is from 1C32-N004B, which is served by an independent and redundant sensing line. During the latter portion of the scram recovery, the FWLC System was selected to "A" level control, which was receiving a false low water level signal. Consequently, the ,

RFP received a feedwater demand signal and supplied water to the vessel even though actual level was high. The RFP trip system did not function in this event to preclude overfilling the vessel because two of the instruments feeding the two-out-of-three-taken-once logic scheme were sensing the false low level condition.

According to the SPDS graphs, reactor water level peaked at a level of 126 inches above instrument zero. It was estimated that approximately 10,000 gallons of water entered the MSLs for the duration of the overfill condition. A significant quantity of the water most likely vaporized to steam. During this time, the TBVs and the MSL drain valves were open, apparently draining the remaining water to the Main Condenser. Based on the piping configurations of the MSLs, HPCI, and RCIC, and on the postulated fluid flow dynamics, it was concluded that most likely a minimal amount of water entered the HPCI steam supply line and that no water entered the RCIC steam supply line. The HPCI steam supply line is equipped with a steam condensate drain pot that would have drained any water that entered the line.

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. TEXT CONTINUATION FACILITY NAME.(1) DOCKET NUMBER (2)- LER NUMBER (5) PAGE (3)

TEAR SEQ hum REV PLANT E. I. HATCH, UNIT 1 05000321 93 012 00 7 or 9 TEXT

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Prior to restart of the reactor, Ceneral Electric performed an evaluation of the j effects of the water entering the MSLs. Based on this evaluation, no safety concerns existed.

• Ii Reactor pressure was at 985 psig prior to the event. As expected, following the scram, pressure decreased to approximately 757 psig. Following the Main Turbine i trip, the TBVs opened and controlled pressure at approximately.920 psig. y Reactor vessel instrumentation provides monitoring capability of critical vessel )

. parameters and provides the appropriate initiating signals when sensed

. parameters exceed prescribed limits. In this event, an instrument sensing line l depressurized, rendering the instruments served by rhe line incapable of )

accurately monitoring their sensed parameters. The instruments associated with  ;

the sensing line and the affect of the condition on the associated ESF are as j follows:

i 1B21-N080A/B: These reactor water level instruments provide  ;

an actuation signal to RPS and PCIS on a low reactor water j level condition. Depressurization of the instrument sensing line 'i caused these instruments to sense a false low level condition and generate a trip signal, resulting in an RPS and a PCIS actuation.

1B21-N093B: This reactor water level instrument provides.a trip signal to the HPCI System on a high reactor water  !

level condition to preclude overfill of the vessel due to l HPCI injection. The logic for this trip signal is a two-of- j two-taken-once scheme and-is not divisionally redundant.

This design is partly due to the fact that the HPCI System is unique among ESFs in that it is a single train safety  ;

system. As such, the system is not designed to be divisionally redundant. Also, the logic scheme precludes a single spurious signal from causing a trip of the system.

As a consequence, the depressurization of the instrument sensing line caused the transmitter to sense a false low water level condition and, given the logic scheme, would .l '

have prevented a trip of the system on an actual high level condition.

1C32-N004A,C: These level transmitters do not perform an ESF function; they provide a level signal.to level indicators 1B21-R606A and C in the Main Control Room, to the FWLC System (1C32-N004A only), and to the Main Turbine and RFP trip system. The affect of the sensing line failure on this instrumentation was previously discussed. i 1B21-N095A: This instrument is a level transmitter and performs two functions. First, it provides a trip signal to the RCIC System (a non-ESF) on a high reactor water level condition to preclude overfill of the vessel due to RCIC injection. The logic for this trip signal is a two-of-l j

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YEAR SEQ hum REV PLANT E. I. IIAIGI, UNIT 1 05000321 93 0'1 2 00' 8 0F 9  !

TEXT:

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two-taken-once scheme-and is_not divisionally redundant. ,

This' design is_ partly due'to'the fact that.the RCIC System- r is a single. train system.. As such, the system.is'not ,

designed totbe.divisionally redundant. Also, the' logic .l scheme precludes a single spurious signal from causing a

~

i trip of the RCIC System. As a consequence, the depressurization .

of the ' instrument s(7 sing line caused the transmitter: to sense a false low water level condition and, given the.logiefscheme,.  :

would have prevented a trip of the system on an actual high level . n-i condition.

1 The second function of this transmitter,is to provide a permissive signal to.the Automatic Depressurization System  !

on a low reactor water level condition. The sensing line failure in this event would have caused.the permissive  ;

signal to be generated at a higher than required level.  :

Consequently, the failure would not affect the ability of' f ADS to-function in an accident. Furthermore, the other inputs required to initiate.the. system would preclude i

. premature initiation of the system. 1 Based on the above information, it is concluded that this event had no adverse impact on' nuclear safety. This assessment applies to all operating conditions. l J

CORRECTIVE ACTIONS >

. . t The packing for isolation valve 1B21-N093B-IV-1 was re-installed and the packing'

~'

- nut torqued.

The packing nuts for instrument valves in Unit I were checked. Twenty-three packing nuts were found to be less than snug. The nuts were subsequently-tightened.

?

During the next Unit 2 Refueling outage, the[ Unit 2 instrument packing nuts of' ,

the type involved in this event will be checked for proper tightness.

l A walkdown of the HSLs; downstream of:the MSIVs was performed and no signs of

damage were identified.

u

General Electr.ic performed an evaluation of.the water'in the MSLs. '

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. ADDITIONAL INFORMATION '

No systems other than those previously identified ~in this' report were affected.

.by thisl event.

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER.(5) PAGE (3)

YEAR SEQ NUM REV PLANT E. I. HATCH, UNIT 1 05000321 93 012 00 9 0F 9 TEXT-

.One similar event occurred within.the past 2 years in which a pressure perturbation on an instrument sensing line resulted.in.an automatic reactor-scram.' This event was addressed in the LER 50-321/91-17, dated 10/9/91. In-this event, a hand-held instrument fell and struck'a sensing line drain valve

~

, stem handle. The impact of the fall.resulted,in the valve partially. opening and the sensing line completely depressurizing. Corrective actions for this event-included counseling. personnel and issuing a plant-wide directive. These actions had no bearing on the condition of the packing' nut and, therefore, could not have' prevented this event.

Failed Component Information:

Master Parts List Number: 1B21-N093B-IV-1 Manufacturer: Dragon Valve, Inc.

Model Number: 60N

. Type: Instrument Manifold Valve Manufacturer Code: -D232 EIIS System Code: JA EIIS Component Code: ISV

-Reportable to NPRDS: Yes Root Cause Code: X c

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