ML20062M543

Intervenor Exhibit I-MFP-57,consisting of 930513 Rev 0 to NCR-93-EM-N014, AFW Pp 21 Steam Supply,2-FCV-37,Failure to Close

ML20062M543
Person / Time
Site:	Diablo Canyon
Issue date:	08/18/1993
From:	AFFILIATION NOT ASSIGNED
To:
References
OLA-2-I-MFP-057, OLA-2-I-MFP-57, NUDOCS 9401070279
Download: ML20062M543 (27)

Text

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NCR DC2-93-EM-N014 Rev. 00

-Q. c DRAFT:

May 13,.1993 93 DC: 28 F6 00 MANAGEMENT. ' SUMMA.Jl During performance of STP P-6B, 2-FCV-37 failed to close on demand from the control room. A' design basis function of:

this valve is to provide c....tainment isolation..

The NCR was initiated on March 15, 1993-(discovery date). l Root Cause: Procedure deficiency in that Electrical Maintenance Procedure MP E-53.10J, Revision 1 (dated 12/18/89), "Limitorque SMB-00 and SB-00 Valve Operator i Maintenance", did not have sufficient detail to ensure that i the quad rings were properly installed after limitorque l l operator disassembly.  ;

Corrective Action

Electrical Maintenance Procedure MP E--

53.10M, Revision 0, "Limitorque SMB-00.and SB-00. Valve .

Operator Maintenance", was issued on January 22, 1993. -This j maintenance procedure contains detailed steps and a composite assembly drawing for re-assembly.of;the limitorque l

! operators. Therefore, there is no'need fer additional corrective action.  !

This draft dated May 13, 1993 reflects the final NCR per the TRG meeting held on May 13, 1993.

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NCR DC2-93-EM-N014 Rev. 00 "

DRAFT: May 13, 1993 NCR DC2-93-EM-N014 AFW PP 21 STEAM SUPPLY, 2-FCV-37, FAILURE TO CLOSE I. Plant Conditions Unit 2 was in Mode 1 (Power Operation) at-100% power.

II. Description of Event A. Summary:

During performance of STP P-68, " Routine Surveillance Test of Turbine-Driven Auxiliary.

Feedwater Pump", Step 10.26.5, flow control valve 2-FCV-37 was taken to the closed position. The position indicating lights indicated mid-penition (both red and green lights illuminated). Locally, the valve was determined to still be in the "open" position. 2-FCV-37 is the Unit 2 isolation valve for off-steam lead number two from the main steam line and is located outside in the pipe rack.

valve closure was stopped by'the closing torque limit switch. The breaker (52-2H-30) for the motor operator on 2-FCV-37 did not trip open when the attempt was made to cycle the valve.

\ On March 15, 1993 NES Engineering determined that

} with the corrosion on the upper bearing, combined with the degraded stem lubrication, the ability of the valve to close with full flow differential pressure - (DP) conditions is suspect.

n .es B.

Background:

Technical Specification 3.7.1.2 requires for Modes 1, 2 and 3: "At least three steam generator auxiliary feedwater pumps and associated flow l paths shall be operable with: (a) Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate vital busses, and (b)

One steam turbine-driven auxiliary feedwater pump capable of being powered from two OPERABLE and redundant steam supply sources."

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o NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993 PSRC TS 3.7.1.2b Interpretation 89-04, dated 5/4/89, requests: "The LCO requires that the Turbine Driven Auxiliary Feedwater' pump must be capable of being supplied from an OPERABLE steam supply system, but provides no guidance on what is necessary for the steam supply system to be OPERABLE." The PSRC Interpretation clarifies:

"For the Turbine Driven Auxiliary Feedwater Pump steam supply system to be OPERABLE, the following must be met.

1. FCVs 37 and 38 must be OPERABLE and open.

2. Check Valves MS-5166 and MS-5167 must be Operable. ,

3. Steam Traps TRP 104, 105, and 106 must be l Operable or bypassed to ensure adequate line condensate removal."

Technical Specification 3.7.1.2 ACTION requires:

"With one auxiliary feedwater pump inoperable, l restore the required auxiliary feedwater pumps to l OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least i HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT. I SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."  !

l C. Event

Description:

i On January 31, 1993 during performance of STP P- I I

6B, Step 10.26.5, flow control' valve 2-FCV-37 was-taken to the closed position. The position indicating lights indicated mid-position (both red and green lights illuminated). Locally, the valve was determined to still be in the "open" position.

Work Order C0110207 was initiated to investigate the failure of the valve to close on demand. The valve was stroked manually, but experienced difficult operation during a small portion of the closing stroke from the full open position. The valve was then stroked electrically and the torque switch was' observed to be " bouncing". On the third valve stroke, the torque switch opened and the valve stopped at approximately ninety-five percent open. When the stem cover was removed, a large amount of "Lubriplate" lubricant was found pooled in the stem nut / lock nut depression.

Inspection of the stem showed that the stem mwommu m Page 3 of 27

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NCR DC2-93-EM-N014 Rev. 00 .

DRAFT: May 13, 1993 lubrication was marginal. The valve stem was i lubricated with'"Lubriplate" and successfully -

stroked manually four_ times. The valve also was stroked electrically with no indication of(torque-  ;

switch bounce. Grease samples were obtained, and.  !

both the limit switch grease and the. actuator grease showed signs'of separation and evidence of i weeping into the limit switch-compartment.was  :

found (Note: this is net unusual). A subsequent. '

test of the grease sample confirmed 1theLgrease in ,

the actuator to be Type EP-0, per design. The l 1 immediate root cause was determined to be a i sticking-valve stem. The valve was returned to service on' February 1, 1993. A work order was prepared to inspect 2-FCV-37 further during the fifth refueling outage (2RS).

On February 4, 1993 a partial internal actuator inspection was' performed.- After removal of the stem cover and the limit switch compartment-cover, and the spring cartridge cap cover, visual inspection found nothing that: could have -caused' the actuator to fail to close. A detailed component inspection was~ planned for 2RS. [REF:

W/O C0110396]

On February 5, 1993 Quality Evaluation Q0010397-was initiated to address potential quality' i concerns related tc this event.

On February 17, 1993, votes testing.was completed

, and no problems wene noted. [REF:'W/O C01104)5) -

i On March 9, 1993 a manual load cell test was.

performed. The as-found thrust was acceptable. .

[REF: W/O C0109271] '

On March 12, 1993, detailed internal' inspection identified significant particulates,' water and corrosion._ The upper. bearing had_ visible corrosion. Preliminary analysis of the. grease-sample showed that_the grease-hadLforeign material ~ l present (i~.e. dirt, rust, metal shavings, etc...).  ;

Engineering analysis determined that_the January.

31, 1993 condition of 2-FCV-37, with the bearing corrosion and stem grease degradation, may have resulted in inability to meet itsLGeneric Letter-muranvuwu Page 4-of 27 l

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. NCR DC2-93-EM-N014.Rev. 00 ,

DRAFT: May 13, 1993 (GL) 89-10 mispositioning closing thrust requirement. STP P-6B, which 2-FCV-37 failed, tests at approximately 2000 pounds thrust. GL 89- l 10 mispositioning thrust requirement.is approximately 6500. pounds thrust. -(REF: _ W/O C0109271) J On March 15, 1993 NES Engineering determined:that the ability e' *? FCV-3' to.close with full flow differ r pressure (DP).was suspect prior, -

to January 3.. 393'with the buildup of corrosion- {

on the upper bearing combined with.the degraded l stem lubrication. (Reference.A0292330,-Eval'09).

However, the successful manual load test-performed March 9, 1993, demonst' *ed that;the GL'89-10

! thrust requirement pri < ' could be-met with the as-found bearing corn ..

D. Inoperable Structures, Components, or Systems that l Contributed to the Event:

None. .

E. Dates and Approximate Times for Major Occurrences:

1. April 16, 1990 FCV-37 Overhauled under W/O R0058494 ,

2. January 31, 1993. FCV-37 failed.to close during STP P-6B. TS l

3.7.1.2 ente *-d.

l A watch.was

' established to meet the requiremonts of TS - 3.6.3 for containment isolation.

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3. February 01, 1993 . Valve stem cleaned and' lubricated, FCV-37..

returned'to service.

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4. February 5, 1993 QOO10397 initiated.

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NCR DC2-93-EM-N014 Rev. 00 -

DRAFT: May 13, 1993

5. March 9, 1993 GL 89-10 thrust requirement met during manual load cell test.

6. March 12, 1993 Internal inspection 4 found upper bearing  ;

corroded. '

7. March 25, 1993 Engineering determined that prior to 1/31/93, Valve 2-FCV-37 may have been unable to close under full flow DP conditions, and was potentially l reportable.

F. Other Systems or' Secondary Functions Affected:

None.

G. Method of Discovery: \

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PG&E plant personnel, during the performance of a scheduled STP P-6B, " Routine Surveillance Test of l Turbine-Driven Auxiliary Feedwater Pump", l identified the problem.

H. Operator Actions: I 1

2-FCV-37 was declared inoperable and TS 3.7.1.2 \

and TS 3.6.3 were entered. 2-FCV-37, cfter "

corrective maintenance, was declared operable-prior to exceeding the LCO action statements.

I. Safety System Responses:

l None.

l l III. Cause of the Event A. Immediate Cause:

l Valve 2-FCV-37 was not capable of closing under full flow differential pressure conditions due__to corrosion of the upper bearing.

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o NCR DC2-93-EM-N014 Rev. 00

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DRAFT: May 13, 1993 B. Root Cause:

Procedure deficiency in that Electrical Maintenance Procedure MP E-53.10J, Revision 1 (dated 12/18/89), "Limitorque SMB-00 and SB-00 Valve operator Maintenance", did not have sufficient detail to ensure that the quad rings were properly installed after limitorque operator disassembly.

C. Contributory cause:

None.

IV. Analysis of the Event A, Safety Analysis:

The Auxiliary Feedwater System (AFWS) serves as a backup supply of feedwater to the secondary side of the steam generators when the main feedwater system is not available, thereby maintaining the heat sink capabilities of the steam generators l (SGs). As an engineered safety feature.(ESF) system, the AFWS is directly relied upon to prevent core damage and system overpressurization  ;

(release of reactor coolant through the pressurizer power operated relief valves or pressurizer safeties) in the event of transients such as a loss of normal'feedwater or a secondary system pipe rupture, and to provide a maans fer plant cooldown following any plant transient.

Auxiliary feed pumps (two 4kV motor-driven pumps and one steam turbine-driven pump) are provided and designed to ensure complete reactor decay heat-removal under all conditions including loss of power and loss of the normal heat sink (the condenser circulating water), while maintaining.

minimum water levels within the steam generators.

The design basis for the AFWS is to ensure that the minimum required flow- (440 gpm) will be delivered to the minimum number of SGs (two SGs),

within one minute, during any. design bases' event.

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l NCR DC2-93-EM-N014 Rev. 00' * '

DRAFT: May 13, 1993 l

The reactor plant conditions that impose safety-related performance requirements on the AFWS are as follows: (1) Loss of main feedwater1 transient, .

(2) Secondary system pipe ruptures,.(3) Loss of .

!. all ac power (station blackout), (4) Loss-of- '

coolant accident (LOCA), and (5) Cooldown. 1 The steam turbine-driven AFW pump supply' train consists of a full-capacity turbine-driven pump which is sized to provide. a' minimum . flow of 930-gpm. The turbine-driven' pump is powered by steam i supplied through two1 full capacity redundant lines taken from two'of the four main-steam lines upstream of the main. steam isolation valves. The redundant supply lines ensure continued .

availability.of steam to the turbine in'the case  :

of a faulted SG or ruptured main steam line- ')

associated with one of 'the supply: lines. -However, there are specific requirements for: component availability and status for the steam supply to ,

the turbine-driven pump to be: considered operable. '

An operable steam supply system for'the AFWS is .I defined as follows: (Modes 1, 2, and 3)

1. Two steam supply lines each with an operable {

remote manual motor-operated isolation valve  !

in the open position (FCV-37 & FCV-38),

2. An operable check valve in each of these lines,

3. Associated steam traps must be operable or

.. -bypassed to ensure adequate conden' sate --

removal, and

4. A single operable motor-operated flowfcontrol valve on the common turbine steam supply line must be capable of automatic-actuation'from closed to open for all applicable ESF.and j manual actuation signals (FCV-95).: l For Modes 4, 5,'and 6 the.AFWS isLnot required by Technical Specification.to-beLoperational. ~

i The two steam. supply lines to'the turbine-driven pump each contain a check valve and.a normally open motor-operated flow control valve-(FCV-37 and FCV-38). The check valves are provided'for passive backflow isolation ofJeither steam-supply 93nctsr93DtNOl4 KM N Page 8 of.27

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i NCR DC2-93-EM-N014 Rev. 00

! DRAFT: May 13, 1993 line in the event of a main steam line-break, allowing continued turbine-driven pump operation using the unaffected steam supply line. The two steam lines join together into a common line prior =

to entering the turbine-driver. This common line contains a normally-closed 125V DC powered motor operated valve (FCV-95). During normal plant operation, the AFWS steam supply lines are pressurized up to '51s flow control valve.

FCV-37 and FCV-38 are remote manual containment isolation valves for the main steam system to the AFW turbine-driven pump, and are Design Class I,

" Group D". .To meet the " Group D" containment piping isolation classification, a single manually operated stop valve is required. This requirement can either be met througn remote manual operation from the control room, or by local manual manipulation if the valve motor or controls have failed. These valves are designed to remain open during operation and close on demand against 1150 l psi maximum differential pressure. These valves fail "as-is" upon a loss of either control or actuator power supply. The valve operators for FCV-37 and 38 are Instrument Class IA since the j valves may be called upon for remote manual operation to isolate a faulted steam generator if required.

None of these reactor plant conditions described above (Loss of MFW, secondary pipe ruptures, station blackout, LOCA, and cooldown) -is cifected  :

by the failure of 2-FCV-37 te close on demand since this valve is normally open and fails "as-is". For these reactor conditions, 2-FCV-37 needs to remain open so as not to impact availability of the turbine-driven AFW pump. l

For the rupture of a main steam line, FSAR Section i 15.4.2.1 states that the MSIV's fully close within ten seconds. Furthermore, for any break,-in any location, no more than one steam generator would blow down even if one of the MSIV's fails to close. This implies that a break in the line common to FCV-95, and FCV-37/38, in conjunction with a the degraded condition of FCV-37 and a single failure FCV-95 could result in simultaneous wrar*nmou Kwk Page 9 of 27 4

NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993 -

blowdown of two steam generators. However, an analysis by Westinghouse has determined that these valves are not required to isolate a break-downstream of these1 valves and upstream of FCV-95.  ;

This is because such-a line-break does not initiate a plant trip and therefore, main, feedwater can be.used to support continued plant' i operation until the line break can be-isolated manually (Reference EOI-8018 and 8062).

Furthermore, engineering has reviewed the associated piping stress calculations-and verified-that the_most probable break point is just i upstream of FCV-95, as assumed in'the Westinghouse analysis.

i The only other safety-related function' associated with FCV-37 is to help mitigate.the consequeices of a steam generator tube rupture (SGTR) accident.

The Westinghouse SGTR analysis assumes FCV-95 is used to stop the steam. driven AFW turbine driven pump to avoid overfill. It is assumed that in the  :

event that FCV-95 fails to~close, flow to the j steam generator would be stopped by an unspecified' ,

means (i.e., by closure of the turbine driven pump- 1 LCVs, closure of FCV-37 and FCV-38, or by tripping l closed FCV-152). ,

Emergency Procedure E-3 addresses operator action during a SGTR. However, discovery of the rupture-is assumed to be preceded by a-reactorLtrip, ,

caused by the difference in temperatureJbetween '

the hot leg and the cold leg (over temperature so-Delta-T), which is covered by. Procedure E-0. Step  !

14 of E-0 directs operator. attention to-the AFW system status. Initially, the operator .;

establishes that the system is delivering at least  ;

470 gpm. If operation of.thelAFWJsystem cannot be j ectablished, the operator is referred to EP FR-H.1  ;

" Response to Loss of Secondary. Cooling." If the  ;

flow requirements are met, the operator's. i attention is focused toward maintaining level in. l the steam generator by placing the motor driven AFW pumps in auto which' controls automatically on level and manually operating the turbine driven i

(TD) AFW pump (step 14.b) to maintain level.

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• NCR DC2-93-EM-N014:Rev. 00-DRAFT: May 13, 1993=

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In step 14.c (EP E-0), when'the SG level reaches 8% in 3 of the 4. steam generators, the operator is' i directed to stop the'TD AFW pump, otherwise ,

control of level. continues by' manual operation of-  ;

the TD AFW pump LCVs. ,Once the decision is made ]

to stop the TD AFW pump, the_ operator has the i option of closing FCV-95, closing FCVs-37.and 38,  ;

closing the TD. pump LCVs (which can be closed by:

remote-manual. action),for by tripping FCV-152 (the turbine trip valve).by local manual' action. 1 Procedure E-0 does not specify which of these- i

'l options to use. . As noted above, the Westinghouse analysis assumes FCV-95 is closed. This ,

assumption is consistent with current operator i practice. ,

Should FCV-95 fail to close, theLWestinghouse analysis assumed the operator would stop' flow.to the steam generators within two minutes. The- -!

means of accomplishing.this.within the two minute ]'

limit was not specified. Current. operator .

practice is to remotely close both FCV-37 and FCV- >

38. Should either of these valves fail to close by remote manual operation, the TD AFW LCVslare ,

closed by remote manual operation, which-can.be accomplished within the allotted two minutes.  ;

Local manual closure of FCV-37 and 38, as well as- )

i local tripping of FCV-152, wouldirequire longer than two minutes. ,

f  !

i Because. closure of the TD AFW^ pump LCVs would' l prevent steam generator overfill,.there-would be L

adequate time to permit either local manual I closure of FCV-37/38 or, FCV-152 tc/ stop the TD AFW pump. Therefore, remote manual operation of FCV-37 and FCV-38 is not required and. local manual operation is an acceptable minimum design bases for these. valves for the SG overfill event.

An additional design function: off FCV-37. and -FCV-38 is to mitigate-a radiologica1 release from a-SGTR-through isolation of'the ruptured SG.- The operators would proceed through E-0 following the reactor trip to step 261of Emergency Procedure E l which refers the operator to Emergency' Procedure E-3 " Steam Generator Tube Rupture" if secondary side radiation is abnormal. At this point, having: H l

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i NCR DC2-93-EM-N014 Rev. 00

• DRAFT: May 13,'1993 assumed FCV-95 has failed to close, there is no requirement to assume an additional failure of the  ;

10% dump valve for the faulted steam generator. '

Likewise, if the 10% dump has failed, there is no requirement to assume FCV-95 has failed.-

l If the 10% dump valve has failed to close, the safety analysis allows 30 minutes for the plant operator to manually close it. The AFW turbine i

exhaust to atmosphere is isolated by closing '

FCV-95.

If FCV-95 has failed to:close, the 10% dump valve is assumed to have been closed. The flow to.the. I AFW turbine is less than the amount of steam that would be released through a stuck open.10% dump valve. .herefore, in excess of-30'rinutes-is available to isolate the flow'from the AFW turbine. This is ample time for the plant-operator to request local manual closure.of FCV-152 or FCV-37 and FCV-38 per E-3 step 3c.

Therefore, the stuck open 10%-steam dump. dose-release bounds a release through the turbine r

driven AFW pump. ,

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Thus, the feature of remote manual (electrical) closure of FCV-37 and 38 is not requiredito {;

j mitigate the radiological release dueLto a SGTR. '

i Therefore, since these administrative controls (EP j E-0 & EP E-3) are already-in-place, the ability to-

~ mitigate the consequences.of an accident is not. -

increased outside the existing.pl' ant resign basis.

The loss of remote manual (electrical) closure capability of FCV-37 will not stop the turbine-

' driven AFW pump from performing its safety .)

function. This valve is an isolation valve onione '

of two steam supplies to the-turbine-driven pump.

Other than maintaining pressure boundary integrity, the valve serves no actual safety-related function relative toLthe operation of the AFW pump.

Therefore, this event-did not adversely affect.the health and safety of the public.

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I NCR DC2-93-EM-N014 Fev. 00 DRAFT: May 13, 1993 B. Reportability:

1. Reviewed under QAP-15.B and determined to be i non-conforming in accordance with Section 2.1.2.

2. Reviewed under 10 CFR.50.72 and 10 CFR 50.73 per NUREG 1027 7nd determined to be not reportable.

The safety function of FCV-37 was' researched to determine whether remote electrical closure is required for accident mitigation. Based on the " Group D" containment isolation classification (remote manual-closure) and the steps in Emergency Procedures E-0 (close FCV-

95) and E-3 (local closure of FCV-152 if FCV-37 cannot be remotely operated), there is'no design basis requirement for. electrical closure.

3. Reviewed under 10 CFR Part 21 and determined that this problem will not require a 10 CFR 21 report, since (a) it is being evaluated under 10 CFR 50.72/73, and (b) it does not involve defects in vendor-supplied services / spare parts in stock.

4. This problem wil1 not be reported-via an INPO Nuclear Network entry.

5. Reviewed under 10 CFR 50.9 and determined to be not reportable since this event does not have a significant implication for public health and safety or common defense and security.

6. Reviewed under the criteria of AP C-29 requiring the issue and approval of an OE and determined that an OE is not required.

V. Corrective Actions

. A. Immediate Corrective Actions

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NCR DC2-93-EM-N014 Rev 00-DRAFT: May 13,J1993

1. The valve actuator stem was manually.and electrically stroked multiple times, after which the valve operated smoothly. '

2. The actuator was electrically stroked several' times. No signs of torque switch l chatter were noted during operation. 2-FCV-37Ereturned to- i service.

3. Quality Evaluation Q0010397 initiated'to track-the problem investigation:and resolution. -An -

as-found inspection and diagnostic. testing were planned for' performance during 2RS.

4. Stem lube inspections (and lubri'cate as.

necessary) of FCV-37/38:& 95 temporarily l increased to a quarterly frequency. .i (REF: 2-b]- l S. Stem covers of 2-FCV-438/439 removed and -

actuators inspected, since:these valves are also located in the pipe rack. 2-FCV-438 was i acceptable; heavy, flaky, rustLand some-standing water were found in the upper section of the stem cover area for 2-FCV-439, however-  !

no water was found in the gear box and' grease 1 i samples appear normal. (Reference 2d-and 5e)  !

l B. Investigative Actions: j Votes diagnostic testing was performed on 2-FCV-37 during the next STP P-6B performance. This. votes e-test found no unusual characteristics,.and the trace was similar.to the DP-test performed during i

2R4. Reference Work Order C00110455. I i

1. Review the overhaul records for valves'l-FCV-37, 1-FCV-438, 1-FCV-439, 2-FCV-37,-2-FCV-438 and 2-FCV-439 to provide assurance:that the

. proper quad rings were staged;and used for valve reassembly.

R'ESPONSIBILITY: C. Shortt DEPARTMENT: PGEM Tracking AR: A0298496, AE'#01' J STATUS: Return l

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' NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993

2. Inspect 1-FCV-37, 1-FCV-438, and 1-FCV-439 for signs of water intrusion, grease degradation, and corrosion.

RESPONSIBILITY: M. Frauenheim DEPARTMENT: PGMB i

Tracking AR: A0298496, AE #02-STATUS: Return l

3. Perform a root cause analysis. Problem statement, as agreed by the TRG is as follows:

" Valve 2-FCV-37 outside of its design basis l prior to January 31, 1993." Subsequent-i

' investigation determined that 2-FCV-37 was not.

outside of design basis. This AE tracks root cause analysis performance only.

RESPONSIBILITY: C. Shortt DEPARTMENT: PGEM Tracking AR: A0298496, AE #03 STATUS: Return

4. Investigate and evaluate the condition of the grease for the upper bearing and the gear box for 2-FCV-37.

RESPONSIBILITY: M. Frauenheim DEPARTMENT: PGMP Tracking AR: A0298496, AE #04 STATUS: Return

5. Determine, if possible, how the wrong quad ring was utilized during the reassembly of valve 2-FCV-37 after its overhaul in 2R3.

RESPONSIBILITY: M Frauenheim DEPARTMENT: PGMB Tracking AR: A0298496, AE #05 STATUS: Return

6. Determine if the apprentice, during 2R3, was qualified to work on limitorque operators.

RESPONSIBILITY: C. Shortt DEPARTMENT: PGEM Tracking AR: A0298496, AE #06 STATUS: Return onmpurmouxwn Page 15 of 27

7 NCR DC2-93-EM-N014 Rev. 00 -

DRAFT

May 13, 1993 4

7. Document basis for non-reportability.

! RESPONSIBILITY: K. Riches J

DEPARTMENT: PTRC i Tracking AR: A0298496, AE #07 4

l STATUS: Return C. Corrective Actions to Prevent Recurrence:

4 i

Electrical Maintenance Procedure MP E-53.10M, Revision 0, "Limitorque SMB-00 and SB-00 Valve Operator Maintenance", was issued on January 22, 1993. This maintenance procedure contains detailed steps and a composite assembly drawing for re-assembly of the limitorque operators. No further corrective actions are required.

MP E-53.10M will supersede MP E-53.10J when MP E-53.10J is rescinded. Rescission of MP E-53.10J is tracked on A0298496, AE #08.

D. Prudent Actions (not required for NCR. closure)

1. Provide protective covers for the FCV's (2-FCV-37, 438 & 439) located outside in the pipe rack area.

RESPONSIBILITY: M. Frauenheim Tracking AR: A0304019

2. Design basis operability following a failed

"~

STP: If there is.any doubt about the cause of a MOV failing to meet STP requirem 7ts, what is the best way to document the consequences of failure and subsequent acceptability to return to service. Please investigate and implement appropriate actions. Consideration of this issue should also be given to components other than MOV's.

RESPONSIBILITY: H. Phillips ECD: 4/27/94 Tracking AR: A0305252

3. Revise DCM S-3B, " Auxiliary Feedwater System,"

to include an additional safety function for the turbine-driven auxiliary feedwater pump steam supply line. Specifically, this line 93nmr outNm Kn Page 16 of 27

t l

' NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993 must be isolated following a SGTR to preclude an offsite radioactive release'if SG-2 or SG has the ruptured tube (s).' Please include the allowable time for isolation of this line following aLSGTR. As' applicable, also revise-DCM T-15, " Radiation Protection" to include this information.

RESPONSIBILITY

• C. Rhodes _ ECD: 4/27/94 Tracking AR: A03052591

-t VI. Additional Information A. Failed Components:

FCV-37, motor. operated flow control valve operator' >

for the AFW pump steam Supply..

Manufacturer: 'Limitorque-Model No.: SMB-00 B. Previous Similar Events:

NCR DC2-89-OP-N009, "AFW Pump 2-1' Inoperable Due-to FCV-37 Being Shut"; ~this is the NCR that resulted in initiation of Tech' Spec interpretation 89-04. On January 17, 1989, action b. of Tech Spec (TS) 3.7.1.2, " Auxiliary Feedwater System", '

was exceeded when both.the steam. driven-auxiliary.

feedwater (AFW) pump 2-1 and motor driven AFW pump l

2-3 were inoperable for greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with i the unit in Mode 1.- Earlier, AFW pump; 2-3 "as: .

removed from service to allow. maintenance on level control valve LCV-115. AFW? pump'2-l'was made inoperable by removal from service'of one steam supply'to the pump'when FCV-37, a steam supply-isolation valve,.was shut to. allow maintenance'on .

the valve motor operator. . The senior : licensed operator had concluded that this activity,did.not.

render AFW pump 2-11 inoperable based;oni a review of the applicable surveillance _testcprocedures.

The procedures implied.that the pump is operable ,

if the pump can maintain full speed and flow with -

only one operable steam supply. The total time that AFW pump 2-1.was out of service was less'than the action statement time of TS13.7.1.2 action a.

wrurmou xwn Page 17 of 27

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t NCR DC2-93-EM-N014 Rev. 00 \

DRAFT: May 13, 1993 The immediate cause of this event is that the senior licensed operator who evaluated.the operability concerns associated with taking-FCV-37 out of service incorrectly concluded that this  !

action would not render AFW. pump 2-1 inoperable. ,

The root cause of this event was a ' lack of:  !

understanding by plant personnel of the design J basis operability requirements of the steam driven AFW pump. This lack of; understanding was 1 due to inadequate' guidance in plant procedures- l which were used to determine the' pump operability. j Applicable plant procedures did not ref1'ect the requirement to have both turbine steam supply paths operable. This requirement is specified in )

the Westinghouse Steam Systems Design Manual ,

(WCAP-7451) and was.not incorporated in applicable plant design bases documentation.

When FCV-37 is closed and'AFW pump 2-1 is'not declared' inoperable, a feedwater line break associated with a failure of AFW pump 2-2.causes i the AFW system to be incapable of delivering _the  !

design AFW flow. If this scenario would have occurred, Emergency Operating Procedure (EP) F-0,

" Critical Safety Function. Status Trees",.would direct operators to EP.FR-H.1, " Response to Loss l of Secondary Heat Sink " This procedure would l instruct the operators to restore at least-460 gpm of feedwater flow to the SGs by performing. local manual valve alignments as necessary to achieve the minimum flow requirements. . The operators would open the closed steam supply toLthn. steam am l driven AFW pump and/or cross-tie the motor driven AFW pumps to establish the required feedwater: flow. {

conditions. . Westinghouse has. performed a feedline '

break evaluation for the past operation.with one i

valve closed on one line of the steam supply to the turbine auxiliary feedwater pump. The following assumptions were made in'this analysis:

j One of the two parallel valves to the auxiliary feedwater pump turbine is-closed for maintenance.

l The main feedline break occurs in the steam l generator feeding the' operating steam-supply, l closing off the'only other path for steam to l i mar 9mm xwn Page 18 of 27 l

l i L i

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NCR DC2-93-EM-N014 Rev.:00 DRAFT: May 13,'1993 t! turbine driven pump. The turbine driven pump - is, therefore,- disabled. l A single failure of the. motor driven pump which is not associated with the faulted SG occurs.

Ten minutes after reactor trip, operator action supplies water to one intact SG by.

isolating the feedline break. Estimated. flow _ '

to the intact.SGLi.s 325 gpm.;

Thirty minutes afterfreactor trip,. operator-action increases the' auxiliary'feedwater to.

440 gpm. This additional feedwater is fed'to at least two steam generators.-

This analysis was performed'for-with powertand without power cases. The results'were shown:to be within FSAR' limits by showing that no boiling occurred in the hot-leg'of the RCS and.that the pressurizer did not fill.

Corrective Actions to Prevent Recurrence include; (1) A revision to the FSAR Update will.be made to  ;

clearly state both main steam supply valves must-be open for the steam driven AFW pump to be operable, and (2) PG&E will. evaluate the need for a technical specification ~ change to clarify that both steam leads-are necessary.in ordernfor.the steam driven AFW pump to be operable.

The corrective actions implemented due to this NCR

! resulted in proper operator actions to declare i FCV-37, conservatively,-inoperable and investigate, i the closing problem with the' motor operator. < J l

C. Operating Experience Review:-

l  :

l 1. NPRDS: I 1

Not applicable.  :

2. NRC Information Notices,-Bulletins, Generic-Letters:

None.

o wur tuson xwn Page 19 of 27 I-

?

k NCR DC2-93-EM-N014 Rev. 00 -

j DRAFT: May 13,.1993

3. INPO SOERs and SERs:

None.

D. Trend Code:

Responsible department EM, and cause code B2.

E. Corrective Action Tracking:

1. The tracking action reg'uest is A0298496.

2. Are the corrective actions outage related?

NO.

F. Footnotes and Special Comments:

1. Even though FCV-37 is not environmentally qualified, the generic question was asked:

"What is the impact on EQ Limitorque operators should the quad rings be accidentally left l out?". For FCV-37, since-it is located l outside in the pipe rack, and'it was exposed I

to large amounts of rainfall, there is a high t

probability of flooding in the actuator housing if missing the upper quad rings. For i the EQ Limitorque's, the protective enclosures l l

or buildings in which they are located, i protect the actuators from inclement weather

^-

(heavy rain fall). In addition,'the EQ Limitorque's are all maintained at -_high _'

level of assurance for proper functionality.

Therefore, if the quad rings were missing form j

an EQ Limitorque, the probability of failure is negligible.

Reference:

AR A0292330, Eval 08, 2.

j Condensation from 10% steam dump valve, _PCV-20, has been seen to splash onto FCV-37.

G.

References:

1. Technical Specification 3/4.7.1.2 and TS PSRC Interpretation 89-04.

a 9hr nmwu wwk Page 20 of 27 1 +

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NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993

2. Related Action Requests.

a. A0292s30; Initiating Action Request. Also refer to associated QE Q0010397.

b. A0295133; Increase stem lube inspection frequency for FCV-37/38/95.

c. A0297566; ?-"pect FCV-438 & 439 for water contanination (also located in pipe rack).

d. A0298244; Water found above housing cover of 2-FCV-439.

3. Licensee Event Report (LER) or other reporting reference.

4. STP P-68, Rev. 26; " Routine Surveillance Test of Turbine-Driven Auxiliary Feedwater Pump."

5. Problem Investigation Work Orders, completion remarks, as applicable.

a. W/O C0110207. "MS-2-FCV-37: Investigate Failure to Operate."

b. W/O C0110396. "2-FCV-37: Perform' Internal Visual Inspection".

c. W/O C0110455. "MS-2-FCV-37: Perform Votes Test". '

d. W/O C0109271, 2R5 detailed inspection.

e. W/O C0112151, 2RS FCV-439 inspection

6. Component Data and Component History (PIMS).

1

7. R0058494 dated 4/16/90 and version of j Procedure MP E-53.10J, Revision 1 dated 1 12/18/89, that was in effect during the 2R3 l actuator overhaul.  !

i

8. Procedure MP E-53.10M, "Limitorque SMB-00 and SB-00 Valve Operator Maintenance," Revision O.

dated 1/22/93; Components Reassembly instructions.

9. DCM S-3B, Revision No. 1; Auxiliary Feedwater l System. Applicable' sections only.

9)ncr*r'93EMN014 KW H Page 21 of 27 l

e NCR DC2-93-EM-N014 Rev. 00

• DRAFT: May 13, 1993

10. Applicable sections of the FSAR Update

a. Section 6.5, " Auxiliary Feedwater System"

b. Section 10.3, " Main Steam System"

c. Section 15.4, " Accident Analyses, Condition IV - Limiting Faults"

11. h.R DC2-89-OP-N009, "AFW Pump 2-1 Inoperable Due to FCV-37 Being Shut".

12. Herguth Laboratories grease analysis reports.

13. Chron 207103; Memorandum from NES to Electrical Maintenance regarding the design basis for FCV-37 and FCV-38.

H. TRG Meeting Minutes:

On March 23, 1993, the TRG convened and considered the following:

l

1. Investigative actions 1-5 were assigned.

j

2. The chronology of the event, as presented in this write-up'was discussed.

3. Problem statement development The upper bearing had visible corrosion. The preliminary analysis of the grease sample showed that the grease had material present (i.e. dirt, rust, metal shavings, etc...).

Engineering analysis had determined'tb?t the .c ~

1/31/93 condition, with the cearing corrosion and stem grease degradation, would not be'able to meet its Generic Letter 89-10 design basis f

I function. This is reportable under 50.73.

STP P-6B, which 2-FCV-37 failed, tests at approximately 2000 pounds thrust. GL-89-10 mispositioning thrust requirement is approximately 6500 pounds thrust.

i

(

TRG PROBLEM STATEMENT: Valve 2-FCV-37 was

{

outside 31, 1993.

of its' design basis prior to January ,

wmruremuku Page 22 of 27'

l l .

NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993 NOTE: Subsequent investigation indicates that the valve is not outside its design basis, but was not operable for a period of time, assumed l to be greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, prior to January 31, 1993. In-operability of 2-FCV-37 also makes the turbine-driven auxiliary feedwater pump inoperable. The limiting conditions of operation for Tech spec 3.7.1.2 were exceeded.

The motor breaker for FCV-37 did not trip open when the attempt was made to cycle the valve f

during the performance of STP P-6B.

In addition, the TRG agreed that the issue, l Whrt should DCPP do in the future when a MOV O ils a STP, needs to be addressed. For r

e.cample, what actions-and associated actions need to be taken if a MOV fails its STP, I corrective maintenance is performed, and then l

the valve passes the STP? The issue is that

the design basis operability needs to be l addressed in a timely manner (i.e. the right l

questions need to be asked). One suggestion l is that a POA could be used.

l On March 30, 1993, the TRG convened and considered the following: 1 The root cause was discussed, and the l following preliminary root cause was 9 presented: Procedure deficiency in that MP E- ,

53.10J did not have enough detail to ensure  !

that the quad rings were installed during l valve assembly.  !

Investigative Action #6.was assigned.

Reportability was discussed. Can this event  !

be assumed to occur at the. time of STP i performance, or since (1) engineering l determined that the corrosion present would prevent the valve from closing during full flow differential pressure, (2) the corrosion is not a short term failure mechanism, and (3) the valve was last overhauled during 2R3 when-l the quad rings were left out; there is __

sufficient evidence to determine that the mmrer.uwnxwn Page 23 of 27

~ _ . .

4 NCR DC2-93-EM-N014 Rev. 00 .

DRAFT: May 13, 1993.

valve was inoperable for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the performance of the STP.

The TRG will reconvene on April 9 to discuss contributory causes and associated corrective actions.

On April 2 and again on April 9, 1993 the TRG convened and considered the following:

1. Design Basis of FCV-37; there is no indication that remote closure of FCV-37 is required for operability of the steam-driven AFW pump. Preliminary presentation of this clarification to the PSRC indicates that the PSRC concurs with this, but is concerned whethel FCV is required for any other accident mitigation scenarios, i.e. SGTR. PSRC requested further investigation.

2. Discussion as to whether there is firm evidence that FCV was inoperable prior to the performance of STP P-6B on 1/31/93. System Engineering' presented the following information to indicate that " firm" evidence {

does not exist to justify assuming outside of  :

the guidelines of NUREG 1022, Supplement 1 '

that the failure did not occur at the time of discovery:

FCV-37 is only required.to be open for )

'*~

steam driven AFW pump functionality. This is substantiated by STP V-3R6 nc* testing the remote operation of the valve from the control room. 'The as-found condition of FCV-37 on 1/31/93, the valve was capable  ;

of being opened electrically, and therefore, the steam-driven AFW pump was never inoperable.

GL 89-10 test for FCV-37 was successfully completed during.2R4.

STP P-6B was successfully completed just 14 days prior to the failure of FCV-37 on 1/31/93.

l e.iam r eurwu m Page 24 of 27 l

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NCR DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993

• The Unit trip on 1/30/93 resulted FCV-37 being exposed to steam. dump condensation moisture, and therefore, this is the most probable cause of the moisture that resulted in final corrosion inoperability of FCV-37.

• Corrosion is a steady degradation, not a step degrade' ion. This is analogous to a pump performance degradation and associated STP performance. NUREG 1022 allows first STP performance failure as the discovery date and does not require-plotting a time-line backwards to try and pinpoint the exact point of failure.

• Since the quad ring is not required for-valve operation, this can be considered as a random failure during the STP performance.

• On March 9, 1993 the valve was tested to determine the as-found thrust of the operator. The thrust was sufficient to meet the design requirements of FCV-37, including GL 89-10 mispositioning thrust valves, even with the corrosion present.

NOTE: A conference call held on 4/13/93 between Engineering, Electrical Maintenance, Quality Assurance, System Enaineering, and l Regulatory Compliance determined that th***e is J no denign basis requirement for remote  !

operation of FCV-37 since the emergency I operating procedures adequately prevent an additional release, for a SGTR with a stLck l open 10% steam dump, path through the turbine -

l driven AFW pump. Therefore, this event does l not appear to be reportable.

April 9, 1993~TRG: Preliminary results of the grease analysis was discussed. . Two grease samples from 2-FCV-37 were analyzed:

l 1. Grease sample from the main gearbox (should be Exxon Nebula EP-0). Visual inspection shows the sample to match the mur muwa m Page 25 of 27

4 NCR DC2-93-EM-N014 Rev. 00 '

DRAFT: May 13, 1993 color of EP-0, but some water and dark swirls were evident, indicating possible contamination. Lab results confirm the grease to be EP-0. Lab results also indicate high silicon, iron and sodium l levels indicating water and dirt contamination. The grease had softened to the lower range of acceptability, but would still be usable as a lubricant.

2. Grease sample from the upper roller bearing area. Sample was completely black with excessive amounts of large metal particles and water. The sample was abnormal and unacceptable for use as a lubricant.

The TRG will reconvene on April 22 to discuss contributory causes and associated corrective actions.

On April 22, 1993 the TRG re-convened and considered the following:

The grease sample preliminary analysis from the April 9, 1993 TRG was finalized. The gearbox grease was acceptable. The upper roller bearing grease was confirmed to NOT'be Lubriplate nor Nebula'EF-0, and the grease was unacceptable for use as a lubricant.

An investigative action to document th basis for non-reportability was agreed to. This AE will be used to track PSRC approval of the-FCV-37 operability requirements (i.e. via a revision to TS Interpretation 89-04).

Design basis operability following a failed STP was discussed. 'It was' agreed that better documentation of investigations is required.

A prudent action was agreed to for the following issue: If there is any doubt about the cause of a MOV failing to meet STP requirements, what is the best.way to document the consequences of failure and subsequent acceptability to return to service.

93ntrwr910.MN014 KWR Page 26 of 27

.' i NCR.DC2-93-EM-N014 Rev. 00 DRAFT: May 13, 1993  ;

The need to revise DCM S-3B, " Auxiliary  !

Feedwater System'," and possibly DCM T-15,

" Radiation Protection" to include an additional. safety function'for the turbine- i driven: auxiliary feedwater pump steam supply line. Specifically, this line must be isolated following a SGTR to preclude an offsite radioactive release'if SG-2 or-SG-3 has the ruptured tube (s). Please' include the  :

allowable time for isolation of this line '

following a SGTR. i A NCR closure date of'7/13/93 was: established.  !

The TRG will re-convene on 5/13/93-for-review- i of the NCR writeup and final signoff.

l On May 13, 1993 the TRG re-convened.and considered

! the following:

The safety analysis, as revised based cn1 NES engineering input, was reviewed. See above. ]

The TRG concurred that the NCR does:not need l l to track.PSRC review of a rescission-to TS-j Interpretation 89-04. This is-reflected in i the response to Investigative Action #7- J l (A0298496, AE #7). I I. Remarks:

None.

l 1

93nemr 93tusoi4 Kwu page 27 of 27 I

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