05000265/LER-1994-010, Forwards LER 94-010 Re Unplanned Scram from Position 48 to 00 During Instrument Maint Surveillance

Forwards LER 94-010 Re Unplanned Scram from Position 48 to 00 During Instrument Maint Surveillance

ML20087F432
Person / Time
Site:	Quad Cities
Issue date:	08/08/1995
From:	Pearce L COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LWP-95-074, LWP-95-74, NUDOCS 9508150280
Download: ML20087F432 (4)

LER-2094-010, Forwards LER 94-010 Re Unplanned Scram from Position 48 to 00 During Instrument Maint Surveillance

Event date:
Report date:
2652094010R00 - NRC Website

text

_

commonwcano w~m comrun>

L QuJd CitiO (etnefJling StJtion 22"Io 206rh Mcnue Nonh Cordova 11. 612 42-9~ so

'Icl M n4A v12il j .

t- LWP 95-074 August 8, 1995 L

l U.S. Nuclear Regulatory Commission  ;

L Document Control Desk Washington, D.C. 20555

Reference:

Quad Cities Nuclear Station Docket Number 50-265, DPR-30, Unit Two

Subject:

Licensee Event Report (LER) 265/94-010 Supplemental Information. l As stated in LER 265/94-010, supplemental information is being provided and is enclosed as Attachment 1. This information constitutes revision 01 to the original LER documentation.

l Attachment 2 is a reproduction of the original text of LER 265/94-010. i This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(iv), "The licensee shall report any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature." '

)

I If there are any questions or comments concerning this letter, please 1 refer them to Nick Chrissotimos, Regulatory Assurance Administrator at 309-654-2241, ext. 3100.

Respectfully, COMMONWEALTH EDIS0N COMPANY QUAD CITIES NUCLEAR STATION

. b. &

L.W. Pearce Station Manager Attachment 1- LER Supplemental Information Attachment 2- LER 265/94-010 (copy) cc: J. Schrage C. Miller INP0 Records Center NRC Region III 500Gd g smcaso7495.twr g 1

9508150280 950808 '

mm..m o,m,mn, DR ADOCK 0500 5 B

- ~ . - ---. -- -. .-. .. -. . - _ - .___ _ .-

ATTACHMENT 1

' l Quad Cities Station - System Engineering Department To: L.W. Pearce Prepared bys Daniel Someter PJS 8-8-35 I CRD System Engineer Froms Paul Aitken $ Reviewed by: Jason Smith LeadNuclear1Bagineer 8N L

Sys Eng Supervie 1

Subject:

NTS ltem #265180H010031ssue A Supplemental Report When -

the Results of the GE Analysis are received.

General Electric performed an analysis of two batches of Scram Solenoid Pilot Valve (SSPV) diaphrgas. Dese h*han were both manufactured in 1989, with one set having been in service since 1990 and the other set having been in service since 1992.

De testing results of the 1990 batch are as follows. He oxidative stability of these diaphragms was near the and of life ney degrade under temperature conditions of 169' C, compared to " good" diaphrases that would degrade at greater than 200* C. De spring constant of the Buna-N material for these di.,, .;s, was not checked. His is because this batch had already developed hardening and cracks. Derefore, no useful information could be gained from this testing.

De testing results of 1992 batch are as follows. He diaphragms were flexible, but exhibited some localized hardening. De spring constant for these diaphragms was comparable to other diaphragms of similar vintage. De oxidative stability for these diaphragms showed that there was sti*i some remaining service life. De diaphragms did not show degradation until 210' C. P De results of this testing showed that the diaphragms that had a service life of greater than three years had degraded enough to affect operability of the control rods. All of the SSPVs that had diaphrages with a service life of greater than three years were replaced with the new Viton elastomer SSPVs during QlF35 and Q2F35. He diaphragms that had been in service since 1992 were left installed until Q2R13, when they were replaced with the new Viton elastomer SSPVs. All of the SSPVs on Unit 2 with the Buna N elastomers have been replaced with the SSPVs with the new Viton elastomers.' All of the control rods on Unit 2 were successfully scram timed. Unit One has 53 HCUs with the old style Buna-N elastomer SSPVs. Dese SSPVs have only been ins service since June of 1994. They will be removed during QlR14, after 1.5 years of service life.~ He analysis has showed that three years of service life is acceptable. Following QlR14, all of the Buna-N elastomer SSPVs will have been replaced in Unit 1.

Dere will be no change in the course of action and commitments indicated by LER 2-94-010 as a result of this testing. His kem can be closed. ,

. _. ~ . . - . - . .- . . . . _ - - - . _ - - - - _ . - _ . - - _ . -, ,

'A.

Commonwrith Edison ATTACHMENT 2 Quad Cites Nucle:r Power Station 22710 206 Avenue North Cordova. Illinois 61242-9740 Telephone 309/654-2241 GGCd4-il9

' September 28,1994 U.S. Nuclear Regulatory Commission ,

Document Control Desk Washington, DC 20555

Reference:

Quad Cities Nuclear Power Station Docket Number 50-265, DPR 30, Unit Two Enclosed is Licensee Event Report (LER)94-010, Revision 00, for Quad Cities Nuclear Power Plant '

Station.

' Ibis report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(iv). The licensee shall report any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature.

The following commitments are being made by this letter:

An expansion of the sampling program of the diaphragms is being performed to gather more data.

Material analysis will be performed by General Electric Co.

An accelerated repair / replacement program of all the suspect SSPV diaphragms on Unit I and 2 began September 28,1994.

A supplement to this report will be issued when the results of the General Electric Co. analysis are received.

If there are any questions or comments concerning this letter, please refer them to Nick Chrissotimos, l Regulatory Assurance Administrator at 309454-2241, ext. 3100.

- Respectfully, COMMONWEALTH EDISON QUAD CITIES NUCLEAR POWER STATION i

(/A__

. G. pbell l N ta:i Manager i GGC/rB/plm Enclosure cc: J. Schrage INPO Records Center C. Miller NRC Region III 4

i

- ~-

.o.

IJCENSEE EVENT REPORT GJE) Peres Rev. 3.0 Fesibey Name (1) Dockas Number (2) rego p) ,

?

Quad Cities Unit Two , 0l5l0l0l0l2l6l5 1 l of l 0 l 5 l Tale (4) l Castrol Red D.11 Usplanned Scram From Poemaa 48 to 00 Dunag Instnuneet M==- Surveillance i meses Dass p) uut Nummer (s) mapan Dess m omerremTeesIm. ewes (s) l Messh . Day Year Year sequensiel Esvimen Meek Day Year FeeHier Daskus Num6ss(s)  !

Nue6er Num6er Nemme j 0l5l0l0l0l l l 0l8 2l9 9l4 9l4 -

0l1l0 - 0 0 0l9 ana marumu B r& , aan r r --1^5r m ai um - .2-2 8 9l4

.as OF Im. a 0l5l0l0'l0l l l s, ^ an a -

l WODE (F) Kneek ens er mese of to Giasmag) 01) 04 an.4me)

~""

ap.4us(e) x so.7s(e)g)0v) 73.713) ruw- 35.405(s)(1)0) 2.36(e)0) -

SD.75(e)G)(v) 75.71(a) l Lavat ap.405(s)o)00 m.M(s)Q) 50.73(a)(2)(vl0 Deer (speauy (10) l9lgy .py8 an.405(ext)010 as.405(a)0)0v) 50.73(a)p)0) 2.73(a)(2)m 50.73(a)GXviiO(A) so.7s(s)g)(ves) m Absmus bei r ens is MM$dc-@ mms , ' 7 A@, -

30.405(n)(1)(v)

N.73(a)p)0E)

$0.75(a)G)(u) Tan)

IJCENSEE s,usi a ACT FOR asaan LER (1: )

N Ah'= . TELarrivns "- l AREA CDDB l xov m a. assoi.iory A , nn. 2789 3 0l9 6 5l4l-l2l2l4l1 couri.ma n owe uNa com sAcu cowo. ant FARERE DESem- r Di , - Raruma @ i cAuss s um ans CImuwan a asAnuracIvm== marum Auta , CAvas saansas comaruna-,a asAnurAwavsma mm,va s Asim f a wrmos m wrnos g i B AlA S l 0 l 1. l Al6l1l0 Y i "

l l l l l l l

)

Q I l l l l l l l l l 1 l~ l pur 'LEhumsaALamrums marm.amar 04) Byssed M Dur Year 2 Sahumanies Q Of yes samphes EXPSCIED SUBbESSION DA12 7 Dass (15) l l l [

As5TBACT Ghat as HOO spesos. i.e.. _.. . flAmen engW gysethe Emm$ GE)  !

I ABSTRACT: j On 08-29-94, Unit 2 was in the run mode at 98% rated core thermal power. The Instrument l Maintenance Department (IM)) was performing a monthly. surveillance when a channel B 1/2 scram signal was intentionally inserted, and caused an unplanned scram of Control Rod Drive (CRD) [AA] D-11 from position 48 to 00.

]

The unplanned scram of CRD D-11 was due to instrument air leakage past the exhaust port )

diaphragm of the 117 scram solenoid pilot valve (SSPV) (SOL). This caused the 127 scram outlet valve to open when the channel B 1/2 scram signal was inserted.

Corrective actions included taking Hydraulic Control Unit (EU) D-11 out-of-service, rebuilding the HCU SSPVs, and rebuilding the 127 scram outlet valve.

LER245\94W10.WPF ei1lil A m .yquF l ~) u // f<f -. - . . - - - _ -. . - . . -.- -

"g'

,. UCENSEE EVDrr REPORT G2R) TEXT CONTINUA 110N Fem Rev. 2.0 Pncn.rrY NAME (1) DOuun NUMBER G) LER NUMBER (6) PAGE (3) e Year seguemmel Revisene

- Number Numdier Quad Cities Unit Two rterh,i.mi ,y- --

-sy.- - e 0l'5l0l0l0l2l6l5

.is ss.si. a. m 9l4 -

0l1l0 -

0l0 2 lOFl 0 l 5 PLANT AND SYSTEM IDENTIFICATION:

G:neral Electric - Boiling Water Reactor - 2511 MWt rated core thermal power.

EVENT IDENTIFICATION: Control Rod D-11 Unplanned Scram from position 48 to 00 during

. Instrument Maintenance Surveillance.

A . C0W ITIONS PRIOR TO EVENT:

Unit: Two Event Date: August 29, 1994 Event Time: 0953 L Reactor Mode: 4 Mode Name: Run Power Level: 98 l

This report was initiated by Licensee Event Report 265\94-010.

RUN (4) - In this position the reactor system pressure is at or above 825 psig, and the reactor protection system is energized, with APRM protection and RBM interlocks in service (excluding the 15% high flux scram).

B. DESCRIPTION 0F EVENT:

At approximately 0953 on 08-29-94, Unit 2 was in the run mode at 98% rated core thermal power. The Instrument Maintenance Department (IMD) was performing Monthly Low and Low-Low Reactor Water Level Analog Trip System Calibration and Functional Testing (QCIS 200-3).

Per procedure, QCIS 200-3, a 1/2 scram signal was inserted on Reactor Protection System (RPS) [JC] channel B. When the channel B 1/2 scram was received, alarm A-3,.

ROD DRIFT, annunciated on the 902-5 panel and an unplanned scram of Control Rod Drive (CRD) [AA] D-Il (14-43) from position.48 to 00 was observed. The channel 8.1/2 scram was reset.at approximately 0954.

An Operator and Shift Foreman (SF) were dispatched to the CRD D-Il Hydraulic Control Unit (HCU) to investigate. The operator checked the scram solenoid pilot valves (SSPV) for air leaks, their associated fuses to determine if they were blown, ensured 4 the solenoids were energized by using a solenoid tester, and measured the scram exhaust line temperature with a thermography gun. No abnormalities were noted; however, a Nuclear Work Request (NWR) deficiency tag was found on the 127 scram outlet valve stating water was leaking by the valve seat. 1 I

Innediate corrective actions included notifying the Unit Supervisor _ (US), Shift i Engineer (SE), a Qualified Nuclear Engineer (QNE), and the System Engineer. The IMD surveillance was also suspended pending re-authorization by the US.

The QNE determined no immediate control rod movements were required to adjust neutron  !

flux profile. ,

UR265\90010.WPF j

LICENSEE EVENT EDORT GEd TEXT CONTINUATION Foem Rev. 2.0 FACILJrY NAME(1) DOuurg NUMBER G) LER NUMBER (6) PAGE 0)

Year sequesmal Revises Number Number Quad Chies Unit Two 0l5lol0lol2l6l5 9l4 -

0l1l0 -

0l0 3 lOFl 0 l 5 i s.a a Energy Induary " '- - systems (EXIE) oodes ese idensi6ed is lhe esas as p0q At approximately 1230, CRD D-11 HCU was~taken out-of-service for Operations and electrically disamed. The 110 surycillance was re-started at approximately 1250.

NWR Q17779 was initiated to rebuild the 117 and 118 SSPVs. Problem Identification Form (PIF) 94-2140 was initiated to investigate the event. An existing work request, Q16701, addressed the repair of the 127 scram outlet valve.

An Emergency Notification System (ENS) notification of this event was completed at 1259 hours0.0146 days <br />0.35 hours <br />0.00208 weeks <br />4.790495e-4 months <br /> on August 29, 1994 to comply with the requirements of 10 CFR 50.72 (b) (2)

(ii).

l There were no systems or components inoperable at the beginning of this event which l could have contributed to this event.

C. CAUSE OF EVENT: l This report is being submitted in accordance with 10 CFR 50.73 (a) (2) (iv), which requires the reporting of any event or condition that results in manual or automatic 4 actuation of any Engineered Safety Feature (ESF) [JE), including the RPS.  !

\

The following is a summary of conclusions and causal Factors (C/F) relating to problems which may have influenced and/or contributed to equipment malfunctions.

C/F: Equipment Specification, Manufacture and Construction The unplanned scram of CRD D-11 was due to instrument air leakage past the exhaust 1 port diaphragm of the 117 SSPV. This caused the 127 scram outlet valve to open when a channel B 1/2 scram signal was inserted.

SSPV 117 and 118 were rebuilt under NWR Q17779. All four diaphrages were hard and brittle; one was found to have a tear approximately 180 degrees around the circumference. Two others had minor defects on the surface. CRD D-Il SSPVs were last rebuilt in March of 1990.

This type of material degradation has been identified in General Electric Company (G080) Rapid Information Comunication Services Information Letter (RICSIL) NO. 069 dated May 2,1994. SSPVs that may contain the suspect diaphragm kits have been identified. The diaphrages from a sample of the identified SSPVs have been examined and showed increased hardness.

Scram outlet valve 127 was rebuilt under NWR Q16701. It was determined the leakage did not contribute to this event.

LER2G9N10 WFF

- .- - -.- ., . .. . _. ~ .~. - _ . .

IJCENSEE EVENT REPORT GJiR) TEXT CONTINUATION Pana Rev. 2.0  ;

[

FACILTIT NAME (1) us mas NUMBliR G) LJiR NUMBER (6) PAGE (3) [

Year seguemmal Revision Nussbar Numihar ,[

a Quad Cities Unit Two 0l5l0l0l0l2l6l5 9l4 -

0l1l0 -

0l0 4 lOFl 0 l 5 sta a Easrgy Induary " ^M sysseenmana) eodes ese idealised in die inst as IXX]

r D. SAFETY ANALYSIS OF EVENT:  ;

1. The safety significance with having.CRD D-11 inadvertently scram is considered minimal. Because the rod scrammed into "00", there were no control. rod blade tip

. enhancement problems.  ;

The scram of D-11 provided negative reactivity. However, if the rod were  !

required to scram on a full-core scram, it would have performed its function.

Therefore, there was.no shutdown margin concern for this event. j

2. This failure mode of diaphragm degradation can also result in 'the failure of a i control rod to scram on an individual scram signal. If a SSPV failure caused the '

failure of a control rod to scram, the backup scram valves would ensure the control rod would insert into the core on a manual or automatic full core scram j signal . >

e There have been two failures to scram since December 3, 1993. Actions have been implemented to expedite replacement of the suspect SSPV diaphrages to reduce j the likelihood of further failures. .

j e All of the 1990 and 1991 vintage suspect SSPVs have recently experienced hot  ;

scram timing. Unit One scram timing was completed 9-2-94. Unit Two scram timing  !

was completed 8-30-94. All of the CRDs with suspect SSPV diaphrages meet the '

Technical Specification scram time requirements.

l o There has been no indication of an adverse trend of the overall Technical  ;

Specification average scram insertion times due to SSPV degradation.  ;

e Based on a sample of diaphrages provided by. Quad Cities, General Electric 1 reports that the hardness of the Quad Cities SSPV diaphrages were not~as severe  !

as that reported by other plants experiencing scram anomalies. General 1 Electric also reports that we should not expect to see a marked increase in the  :

failure rate.

1 E. CORRECTIVE ACTIONS COMPLETED: q J

=The-immediate corrective actions involved the Operations Department sending an 1 operator and Shift Foreman to visually inspect HCU D-11. The operator checked the 'l SSPVs for air leaks, checked for blown fuses, ensured the solenoids were energized, and checked for flow in the scram exhaust line. No abnormalities were noted; however a deficiency tag was found on the 127 scram outlet. valve stating water was leaking by i the valve seat. -l CRD D-11 was taken out-of-service and electrically disanned.

uiR26R94W10.WPF )

-_ -- - ._______-_____________.__-)

4

, UCENSEE EVENT REPORT GRt) TEXT CONTINUA 110N Form Rev. 2.0 FACILTTY NAME 0) DOCKET NUMBER G) L.ER NUMBER (6) PAGE 0)

Year Sequenual Revince Number Number Quad Cities Unit Two 0l5l0l0l0l2l6l5 9l4 -

0l1l0 -

0l0 5 lOFl 0 l 5 TEXT Emergy Indusuy idean6couon Symem (EDS) codee are ideau5ed in the text as [XX)

The SSPVs,117 and 118, were rebuilt under NWR Q17779; this included the replacement of the diaphragas. The scram outlet valve,127, was rebuilt under NWR Q16701. This  ;

work was completed and CRD D-11 was successfully tested and declared operable on 09-04-94.

CORRECTIVE ACTIONS TO BE COMPLETED:

An expansion of the sampling program of the diaphragms is being performed to gather more data. Material analysis will be performed by General Electric Co.

(NTSi 26518094001001).

An accelerated repair / replacement program of all the suspect SSPV diaphrages on Unit I and 2 began September 28, 1994. (NTS# 26518094001002).

A supplement to this report will be issued when the results of the General Electric Co. analysis are received. (NTS# 26518094001003).

F. PREVIOUS EVENTS: ,

General Electric Company RICSIL 069 Rev. I has identified this as an industry wide problem.

The Nuclear Tracking System data base listed one LER involving CRD drift or scram at i Quad Cities Station for the period from January 1, 1989 to present.  !

1 e LER 254/94-001, dated 1-24-94, Rod Drift of Control Rod H-1 From Position 48 to 14 During Instrument Maintenance Surveillance COMPONENT FAILURE DATA- I G.

l This event is reportable to the Nuclear Plant Reliability Data System (NPRDS). l l

The SSPVs are manufactured by the Automatic Switch Company (ASCo) (A610) with manufacturer part number HVA90-405-2J.

The rebuild kits for the ASCo valves are manufacturer part number 204-137.

l l

l IIR265\94210.%TF l

• QCAP 1000-6 UNIT 1(2)

• REVISION O .

~

ATTACHMENT A (Page 1 of i)

OFFSITE REVIEWQuad AND INVESTIGATIVE FUNCTION TRANSMITT Chies Nuclear Power Station Reference Nunh; Date: h///ff Sspsk,,,s L / fe n, d & LER nW'yy- oi o Ssblect: ,, ,

Sshmitted kr. ~1"; sag 8u hel  !

/

PWM.U$i.

p G ,

1. Safety Evolustions,tiQIIrwolving an unreviewed safety question as denned in 10CFR50.50 for.

a. Changes to procedures as descrbed in the Safety Analysis Report.

b. Changes to equipment or systems as described in the Safety Analysis Report.

c. Tests or W E descrbed in the Safety Analysis Report.

2. Proposed changes which Irwolve an unreviewed safety question as defined in* 10CFR50.50. 1

a. Procedure changes. I

b. Equipment or system changes.

- I

c. Tests or ==Iments.  !

3. PE- # changes to the Technical Se or Operating Uoonse.

4. Noncompliance with codes, regidations, orders, Technical SpecNicuitions, license r=M-4 or irtomal pr~~4ures or instructions having nudear safety signNicance.

5. Signficant operating abnormellties or devistions from normal and expected performance of piare equipment that anects nucisar safety.

X 6. M REPORTABt.E EVENTS (LERs only).

7. M recognized indications of an unar*4gused danciency in design or operation of safety.

reisted strutsures, systems, or cor -reres.

8. M changes to the Station Emergency Plan prior to implementation.

9. M tems relened by the Systems Engineering Supervisor, Station Manager, Ste Vice 1

Proeident, and General Manager of Quanty Programs and Asseeements.

~'

• 2*

sMfpigfofgjATiblEYMNSIN4+ '

10. Other OSR leams/ Documents E addressed above.

4 This Transmittal is being made in accordance with Quad Cities Nuclear Power Station

' Technical SpecNlestions 6.1.G.2.d(1) for information only. No specmc action is required unloos deemed necessary by ORsite Review and Irwestigative Function.

s .

s. . - . , ...,,, .-.---. ,. .., _ _ . _ , , . . . - - - , , , . , - . , - . . . , ~ , , , , .