ML17263A330

LER 93-003-00:on 930328,discovered Failures of Safeguard Svc Water Sys During Planned Maint.Caused by Design & Operating Environ.Replaced Affected Valves W/Qualified Spares.W/ 930720 Ltr

ML17263A330
Person / Time
Site:	Ginna
Issue date:	07/20/1993
From:	Backus W ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17263A329	List: ML17263A327 ML17263A330
References
LER-93-003, LER-93-3, NUDOCS 9307280206
Download: ML17263A330 (18)

Text

NRC FORM 366 US. NUCLEAR REGULATORY COMMISSION

~ (64)9) APPROVED 0MB NO. 31504))04 EXPIRES'/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F630). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, ANO TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503, FACILITY NAME (11 DOCKET NUMBER (2) PAGE i na Nuclear Power Plant 0 5 0 0 0 2 4 4 1 OF 0 9 During Planned Maintenance, Failures of Safeguard Service Hater System Here Discovered EVENT DATE 15) LER NUMBER (6) REPORT DATE (7) OTHFA FACILITIES INVOLVED IS)

MONTH IP)y'..'SEOVENTIAL RKVI)ION DAY YEAR YEAR OAY YEAR FACILITYNAMES DOCKET NUMBER(S)

NUMBER NUMBER MONTH 0 5 0 0 0 03 28 3 3 0 0 3 0 070 9 9 3 0 5 0 0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE RLOUIREMENTS OF 10 CF R (): IChect one or more ol the Iollowinpl (11)

MODE (9) 20A02(lr) 20.405(c) 50.73(e l(2) (lvl 73.71(lr)

POWER 20AOS(e) (1) (II SOM(c) (1) 50,73(ele)(2) l(2)(v) 73.71(e)

LFYEL 0 0 0 20AOS(e l(1)(9) 50.36(c) l2) 50.73(e) l2) ivB) OTHER ISpecily in Ahttrect trelow end In Text, iVRC Form 20AOS(el())(ill) 50.73(e l(2)(ll 60.73( ~ ) (2)(vill)(Al 36$ AI 20AOS(el )(lv) 50.73(s)(2) (9) 50.73 (vill l(B) 0 20AOS(el (1)(vl 50.73(s 1 (2) I III)

Voluntary Report 50.73(e) l2) (el LICENSEE CONTACT FOA THIS LER (12)

NAME TELEPHONE NUMBER Hesley H. Backus AREA CODE Technical Assistant to the Operations Manager 3 15 524 44 46 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPOAT (13) e'.6 c CAUSE SYSTEM COMPONENT MANUFAC TURER EPORTABLE TO NPRDS CAUSE SYSTEM COMPONENT MANUFAC- EPORTABLE F~+~ ' I~+.

TURER TO NPRDS B B1 ISV C 84 Y SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTED SUBMISSION YES IIIym. COmplete EXPECTED SU84IISSIOII DA TEI DATE (15)

NO ABSTRACT I(,lmlr to 1400 rpecet, I e., epproxlmetely filteen rlnple-specs rypewritten liner) 116)

On March 28, 1993 at approximately 1200 EST, with the reactor in the defueled condition, the Maintenance Department, during valve improvement program maintenance, dis'covered that two manually operated Service Water System Valves, that were required to be open during normal operation, were failed in the closed position, and some isolation valves had excessive seat leakage.

No immediate operator action was necessary because the failures were identified on out of service sections of the Service Water System.

The cause of these events was determined to be partly due to'esign and partly due to the operating environment. (This event is NUREG-1022 (B) and (X) cause codes).

Corrective action taken was to replace the affected valves with qualified spares. Corrective actions to prevent recurrence are discussed in section (V) (B) of this report.

9307280206 930720 "I(A PDR ADQCN 05000244 S PDR NRC Form 366 (669)

NRC FOAM 366A US. NUCLEAR REGULATOAYCOMMISSION APPROVED OMS NO. 31500(0(

(SJ)9) 5 X PI A ES: 4/30/92 IMATEO SURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENTSAANCH (P430). (J.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. OC 20555, ANO TO 1HE PAPERWORK REDUCTION PROJECT (3(5001041. OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME ll) DOCKET NUMBER (2) L'ER NUMSER (5) PAGE (31 YEAR SEOUENTIAI REVISION gg NUMSER NUMSER R.E. Gonna Nuc1ear Power P1ant os 000244 9 3 0 0 3 0 0 02 OF 0

TEXT ///mare e/reae /e n/JI/urer/ Iree ////ane ~ HRC Farm 355A 3/ ((2)

X aZ-ZVENT Pupa CONDITIONS The plant was in the Cold/Refueling Shutdown mode with the reactor in the defueled condition. Phase Five (5) of the Valve Improvement Program (VIP) was in progress with major emphasis on the Service Water System valves. Normal valve degradation had been observed in the previous 4 phases of the VIP. During Phase Five, a more serious degraded condition was discovered for Crane Model 101XU valves.

The following is a listing of the recent problems exper-ienced with these Crane Model 101XU valves:

0 May 1990: First failure of a Crane Model 101XU valve was identified. This failure was in a non-safety related application, and was documented on Work Request/Trouble Report (WR/TR) g9000910 for Service Water System valve 4675 (Service Water Inlet Isolation Valve to Main Generator Hydrogen Side and Air Side Seal Oil Coolers).

0 April 1991: Indication of a possible second failure of a Crane Model 101XU valve was identified. This possible failure, also in a non-safety related application, was documented on WR/TR f9100754 for Service Water System valve 4690 (Service Water Inlet Block Valve to Turbine Lube Oil Cooler "B").

September 1991:. After several troubleshooting efforts as followup to WR/TR f9100754, radiography of valve 4690, documented on WR/TR f9122140, confirms failure of valve 4690.

0 April 1992: During disassembly and repair of valve 4690, the failure mode is determined to be the same failure mode as for valve 4675.

NRC Form 366A (689)

I

NRC FORM 366A IAS. NUCLEAR REGULATORY COMMISSION (SJ)9) APPROVED OMB NO. 3'1500104 6 XP I R ES: 6/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENTBRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (3(500(04). OFFICE OF MANAGEMENTANO BUDGET, WASHINGTON, OC 20503.

FACILITY NAME (11 DOCKET NUMBER (2) PAGE (3)

LER NUMBER (Sl YEAR SEQUENTIAL n.S" REV (SION g@ NUMSER ~/.( NUMBER R.E. Ginna Nuclear Power Plant TEXT /l/mac opooo /J neo/oN/ ooo ad/6ono/HRC  %%dnn 35643/ (12) 0500024493 0 0 3 0 0 03 OF 0 9 DESCRIPTION OF EVENT A. DATES AND APPROXIMATE TIMES OP MAZOR OCCURRENCES:

o March 28, 1993, 1200 EST: Event date and time.

o March 28, 1993, 1200 EST: Discovery date and time.

EVENT:

On March 28, 1993 at; approximately 1200 EST, with the reactor in the defueled condition, the Maintenance Department was performing Phase Five (5) of the VIP.

As included in the 1993 VIP, Service Water System valve 4669 (Service Water Inlet to Emergency Diesel Generators "A" and "B" Crosstie) (safety related) was to be refurbished and Service Water valve 4738 (Service Water Loop "B" Root Valve to Auxiliary Building Motor Coolers) (safety 'related) was to be replaced. Valve 4738 was planned for replacement, vice refurbishment, due to unavailability of repair parts. When the internals of the valves were exposed, the existing conditions revealed that the valve disk had separated from its valve stem. Both valve disks were found in the closed position with their stems separated from the disk and fully retracted. This failure mode is undetectable under normal operation and with existing routine periodic testing,. due to parallel flow paths. (The normal at power condition for these valves is "locked open").

Also during Phase Five (5) of the VIP, special performance tests identified other Service Water System valves with unexpectedly high leakage past the valve seat with the valve in the closed position.

These valves perform an isolation function, and this leakage degraded their isolation capabilities.

NRC Form 366A (SJI9)

t NRC FORM 368A US. NUCLEAR REGULATORY COMMISSION (689) APPROVED OMB NO. 3)500104 E XP I 8 ES: E/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) I FORMATION COLLECTION REOUESTI 50J) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENTBRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504)04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACII.ITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR SEQVENTIAI. REVISION N VMS E8 NUMSER R.E. Ginna Nuclear Power Plant o so oo24493 0 0 3 0 0 P 40FP 9 TEXT ///more e/Noe /e teqoka/, we af //I/ona///RC FomI 38848/ (17l C. XNOPERABLE STRUCTUE&S F COMPONENTS F OR SYSTEMS THAT CONTRXBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

Crane Model 101XU valves are also installed in the Component Cooling Water (CCW) and Auxiliary Feedwater Systems. See Section (V)(B) for more detail on valves in these systems.

E. METHOD OF DXSCOVERY:

These events were discovered during planned VIP maintenance for the.1993 Annual Outage.

F. OPERATOR ACTION:

As 'these were component failures identified on out of service sections of the Service Water System, no immediate operator action was necessary.

G SAFETY SYSTEM RESPONSES:

None.

XXX CAUSE OR &TENT A. XMMEDXATE CAUSE The immediate cause of valves 4738 and 4669 heing unknowingly in the closed position was due to a separation of the valve disk from the valve stem.

The immediate cause of excessive leakage was due to the general deterioration of isolation valves.

NRC Form 388A (689)

NRC FORM 388A US. NUCLEAR REGULATORY COMMISSION

((W)9) APPROVED 0MB NO. 31504104 E XP I R ES: 4/30/92 LICENSEE E IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS I T REPORT ILER) INFORMATION COLLECTION REQUEST: 508) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME I'I) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

YEAR SEQUENTIAL 3<( REVISION NUMBER Iuy NVM ER R.f. Ginna Nuclear Power Plant TEXT ///Ruuo 4pooo b noduled, uoo eI/cNone/N/IC Foun 3%43/ l)T) osooo24493 0 0 3 0 00 5O" 0 9 B. INTERMEDIATE CAUSE:

The stem and disk of valves 4738 and 4669 had separated due to a variety of corrosion effects.

Isolation valve deterioration was due to a variety of factors, including corrosion, wastage, and environmen-tal conditions, resulting in valves not fully isola-ting.

ROOT CAUSE.

The underlying cause of the corrosion effects on valves 4738 and 4669 was due to the use of dissimilar metals in the manufacture of the stem and disk, combined with prolonged exposure to raw service water and differential aeration cell (concentration cell) corrosion due to stagnant conditions surrounding the tee slot area in the valve bonnet.

The underlying cause of the valves not fully isolating was due to prolonged exposure to the erosive and corrosive effects of raw service water.

ANALYSIS OP EVENT These events are being voluntarily reported using the guidance of NUREG-1022 (Licensee Event Report System), and Supplement 1 to NUREG-1022. While the safety significance of these specific events does not require submittal of a Licensee Event Report, these types of degradation could be safety-significant at other plants, depending on the valve applications. is intended to alert other utilities and theThisNRCreport of problems in applications where corrosion can occur between the valve stem and disk, in raw water applications, and of the potential for degradation of isolation capabilities due to valve deterioration.

These events are related to, but do not meet, the reporting requirements of 10 CFR 50.73, item (s)(2)(v) and (a)(2)(vi);

which requires reporting of conditions, "that alone could have prevented the fulfillment of the safety function",

but where, "individual component failures need not be reported".

NRC Form 388A (889)

NRC FORM366A UB. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31500104 L

EXPIRES: 4/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) INFORMATION COLLECTION REQUEST: 60.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENTBRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504))04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503.

FACILITY NAME (Il DOCKET NUMBER (2)

LER NUMBER (6) PAGE (3)

" 'j SEQUENTIAL '..':jP IIEYIEION YEAR N'UMB E rl N?A rr UM6 6 rl R.E. Ginna Nucl ear Power Pl ant p 5 p p p 2 4 4 9 3 0 0 3 0 0 0 60F 0 9 TEXT ///mare 4/ra>>/4rer/rr/rerL Iree /I/orre/HRC Avm36643)(17)

An assessment was performed considering both the safety consequences and implications of this event with the

.following results and conclusions:

As part of this assessment, an evaluation was performed concerning the Service Water System operability, prior to the 1993 Annual Outage, due to the effects of Service Water System valve leakage from the valves designed to isolate the non-essential service water during an accident with loss of offsite power and due to the effects of the two failed close essential service water cross tie valves.

The evaluation considered 3 accidents (i.e. Containment Integrity, Loss Of Coolant Accident (LOCA) and LOCA Recirculation) using the following assumptions:

o Total service water isolation valve leakage of approximately 1100 gpm, based on a detailed results of .special performance tests conducted during the 1993 outage.

o One service water pump operating.

o Single failure of the "A" Emergency Diesel Generator.

o Loss of offsite power.

Based on the above assumptions the main thrust of the evaluation was to investigate whether the identified valve failures and leakage could have adversely impacted nuclear safety due to changing the service water flow to the critical components for required accident cooling. The critical components considered were the Emergency Diesel Generator Coolers, the Containment Recirculation Fan Cooling Coils, the Containment Recirculation Fan Motor Coolers and during the recirculation phase of the accident, the Component Cooling Water heat exchangers.

Conclusions from the above evaluation indicate that all critical component flows were acceptable.

NRC Form 366A (689)

)~

NAC FOAM 366A US. NUCLEAR REGULATORY COMMISSION (669) APPROVED OMB NO. 31500104 EXPIRES: E/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT ILERI INFORMATION COLLECTION REQUEST: 50A) HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555. AND TO 1HE PAPERWORK REDUCTION PROJECT (31504104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503.

FACILITYNAME (1) DOCKET NUMBER (2)

LER NUMBER (5) PAGE (3)

YEAR SEQUENTIAL REVISION NVM ER NUMBER R.E. Ginna Nuclear Power Plant TEXT //I mac space /t mgv/nnt, IIsP ada//I/one/NRC FnmI 3//549/ (17) osooo24493 0 0 3 0 0 0 7 OF 0 9 The potential for interruption of Service Water flow to the Safety Injection (SI) pump thrust bearings was evalu-ated. This evaluation determined that flow from the redundant Service Water line to the SI pumps was adequate using the assumption outlined above.

Based on the above, it health and safety was assured at all times.

can be concluded that the public's V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

The failed Crane Model 101XU Service Water valves were replaced with qualified spares of a different design and material composition, were tested satis-factorily and were returned to service. Other degraded Service Water valves were also replaced with qualified spares.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE.

As part of the VIP and to prevent recurrence of the Service Water System valve failures, all other Crane Model 101XU valves in the Service Water system were assessed for functionality and those valve warranting replacement were replaced during the 1993 Annual Outage. In addition, selected Crane Model 101XU valves in the CCW and Auxiliary Feedwater Systems were inspected, with satisfactory results.

NRC FonII 366A (6 69)

E I'

NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31500104 EXPIRES: 4/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) INFORMATION COLLECTION REOUEST: 50JI HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS CONTINUATION 'EXT AND REPORTS MANAGEMENTBRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

LER NUMBER IS)

YEAR SEOVENTIAL ??X% REYlSION NUMBER '4?". NVMSEII R.E. GInna Nuclear Power Plant o so oo24493 0 0 3 0 0 0 8 OF 0 TEXT ///mme 4/Joco /4 nqukat, uoo //I/ooo/N/IC Foon 3664'4/ I)7)

As part of the VIP and to prevent recurrence of the Service Water System valve type failures, all remaining Crane Model 101XU valves in the Service Water System are scheduled to be refurbished or replaced during the 1994 Annual Outage. In addition, remaining Crane Model 101XU valves in the CCW and Auxiliary Feedwater Systems will be inspected in 1994.

results warrant refurbishment or replacement, the If the inspection valves will be replaced.

As a result of tests performed on the Service Water System, the scope of maintenance was increased, and other Service Water System valves were also inspected during the 1993 Outage. Valves found to be excessively deteriorated were replaced, and other valves were refurbished, if warranted.

ent/replacement will continue during the 1994 and Inspection/refurbishm-1995 Annual Outages.

Based on the results of the VIP inspection/refur-bishment/replacement, a preventative maintenance frequency, for valves in the Service Water System, will be established as part of the Reliability Centered Maintenance process.

ADDITIONAL INFORMATION FAILED COMPONENTS:

The failed Crane Model 101XU valves were manufactured by Crane Company.

B. PREVIOUS LERs ON SIMILAR LRG94TS:

A similar LER event historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be identified.

NRC Form 366A (669)

NRC FORM 368A US. NUCLEAR REGULATORY COMMISSION (889) APPROVEO OM 6 NO. 3)500)08 6 xpIREs: 8/30/92 IMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E T REPORT (LER) INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (M)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555, AND TO 1ME PAPERWORK REDUCTION PROJECT (31600'l04), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON,OC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

LER NUMBER (8)

YEAR jXB> SEQUENTIAL yN( NVMSSR ~@ NVMSSII R.E. Ginna Nuclear Power Plant TEXT ///mare e/reoe /e mr/rrka/ we aA////one//r/RC Fomr 38683/ ( 17) osooo2443 0 3 0 0 0 9 OF 0 C. SPECIAL COMMENTS:

The industry was informed of these failures via Nuclear Network on April 8, 1993. A report of component failures will be submitted to the NPRDS System.

These failures may also be undetectable at other plants, under normal operation and with existing routine periodic testing, due to parallel flow paths.

Failures at Ginna were only detected during valve disassembly and/or replacement, or as a result of special performance tests. Other utilities may want to consider the benefits of enhanced testing or maintenance evaluations to detect these types of failures.

NRC Form 388A (689)