text

LER-2088-001,
Event date:
Report date:
	3242088001R00 - NRC Website
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LER 88-001-07:on 880102,manual Reactor Scram Occurred Due to Decreasing Main Condenser Vacuum.Reactor Power at 55% & Vacuum Decreased to 22 Inches Mercury.Caused by Leaks on Main Turbine Piping.Piping repaired.W/900801 Ltr
	ML20056A340
Person / Time
Site:	Brunswick
Issue date:	08/01/1990
From:	Harness J, Jones T; CAROLINA POWER & LIGHT CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	BSEP-90-0536, BSEP-90-536, LER-88-001, LER-88-1, NUDOCS 9008070039
	Download: ML20056A340 (11)
	v • d • e

p gpgg, Carolina Power & Light Company Brunswick Nuclear Project P. O. Box 10429 Southport, NC 28461-0429 August 1, 1990 FILE: B09 135100 10CFR50.73 SERIAL: BSEP/90 053,6 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 DOCKET NO. 50 324 LICENSE NO. DPR 62 SUPPLEMENT TO LICENSEE EVENT REPORT 2 88 001 Centlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Supplemental Licensee Event Report is submitted. The original report fulfilled the requirement for a written report within thirty (30) days of a reportable occurrence and was submitted in accordance with the format set forth in NUREC 1022, September 1983.

Very truly yours, eg,hvvww J. L. Harness, Ceneral Manager Brunswick Nuclear Project TMJ/mcg.LTR2 Enclosure cc: Mr. S. D. Ebneter Mr. E. C. Tourigny BSEP NRC Resident Office

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,..c, - . . . . . - ,, - ,nei While performing a routine reactor shutdc,.m in preparation for the Unit 21988 refueling / maintenance outage, a manual reactor protection system trip (scram) was initiated at 0017 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months on 1/2/88, due to a decreasing condenser vacuum.

Reactor power was approximately 55% and vacu'.un had decreased to

, approximately 22 inches mercury. During the expected vessel level shrink

following the scram, vessel level decreased to approximately 153 inches, thus initiating primary containment isolation valve groups 2, 6, and 8 at low level 1 (> 162.5*). Operator verification of these valve closures determined that the group 2 valves 2 016 F003. -F004, F019, and F020 failed to close. These are the inboard and outboard isolation valves for the drywell floor drain sump (F003, F004) and the drywell equipment drain sump (F019, F020). The remaining

safety systems operated as designed during this event.

I Investigation of the decreasing vacuum condition determined it resulted from l

numerous leaks on the main turbine and main steam reheat interconnecting piping to the main turbine, which were repaired during the unit outage. The -

, failure of the group 2 PCIVs to close was caused by the failure of the associated solenoid valves to open due to adherence between the lower brass '

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0 l7 0l3 of I l0 femu smeea mawnn Initial Conditions At approximately 2015 hours0.0233 days 0.56 hours 0.00333 weeks 7.667075e-4 months on 1/1/88, a power reduction was commenced from 69% power to commence a scheduled 16 week refueling and maintenance outage.

This initial condition tapresented the maximum attainable reactor power due to operating cycle fuel depletion. After approximately two hours, at a power level of approximately 50%, problems were encountered maintaining condenser vacuum. At 0017 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months on 1/2/88, a manual reactor protection system (EIIS/JC) trip (scram) was initiated with main condenser (EIIS/SG) vacuum at approximately 22 inches mercury (lig) and decreasing in anticipation of an automatic scram due to the main turbine (EIIS/TA) trip on low vacuum at greater than 304 power. At the initiation of the scram, plant emergency core coolin6 systems and other safety systems were operable.

Event Descriction Ar 2015 hours0.0233 days 0.56 hours 0.00333 weeks 7.667075e-4 months on 1/1/88, reactor power reduction was cominenced in preparation for the scheduled refueling / maintenance outage. As power was decreased, condenser off gas flow slowly began to increase such that, at 2125 hours0.0246 days 0.59 hours 0.00351 weeks 8.085625e-4 months , the augmented off-gas system bypass valve (EIIS/WE/FCV) automatically opened due to high flow (setpoint; 150 scfm). This increase in off gas flow was not unexpected in that it had been noted that off gas flow had increassd during previous power reductions. This off gas power relation is believed to be caused by minor steam leaks at high p wer levels which become vacuum leaks at lower power levels. Although off-gas flow was increasing, condenser vacuum was showing slight improvement. Power was reduced to 48% at 2215 hours0.0256 days 0.615 hours 0.00366 weeks 8.428075e-4 months and 28 steam jet air ejector (SJAE) (EIIS/Sil/EJR) was secured with 2A SJAE remaining in half load per the plant shutdown procedure. After securing 2B SJAE, condenser vacuum began to decrease and the 2A SJAE was placed in. full load at 2245 hours0.026 days 0.624 hours 0.00371 weeks 8.542225e-4 months , and reactor power was increased to 51%. These actions caused-vacuum to reverse the downward trend and start improving. With vacuum improving, efforts were initiated to identify vacuum leaks for repair / isolation to allow the recommencement of the scheduled power reduction.

At 2345 hours0.0271 days 0.651 hours 0.00388 weeks 8.922725e-4 months , vacuum again began to decrease with the decreasing trend being at a higher rate than had been observed during the initial decrease following the securing of 2B SJAE. No evolution had taken place during the previous hour which would have caused this change. Attempts to place the 2B SJAE in service were unsuccesr N due to a low output from the startup permissive temperature instrumem 't;S/JA/TC). Power was increased to approximately 554 by increasing recircule: Da flow and control rod withdrawal in an effort to terminate the decreasing vacuum trend. These efforts appeared to have no effect. With the vacuum continuing to decrease, the decision was made to manually scram the reactor prior to receiving the automatic scram due to the turbine trip on low condenser vacuum at greater than 304 power.

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At 0017 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months on 1/2/88, a manual scram was initiated at a condenser vacuu:s of approximately 22 inches Hg and decreasing. A normal scram recovery was initiated using the emergency operating procedures. Immediately following the ,

scram, reactor vessel level decreased to approximately 153 inches due to expected void collapse and returned to the normal operating level. As a result of decreasing below a reactor vessel level of 162.5 inches (low level 1), an automatic scram signc1 was initiated along with an automatic primary containment isolation system (PCIS) (EIIS/JM) signal for groups 2, 6, and 8.

A verification that these automatic functions, per the emergency operating -

procedures, occurred at 0020 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months determined that the PCIS group 2 valves (EIIS/JM/ISV), 2 016 F003, F004 -F019, and F020, failed to close. These valves are the inboard and outboard isolation valves (both located outside the primary containment) for the drywell floor drain sump (EIIS/IJ/SNK)

(F003, F004) and the drywell equipment drain sump (EIIS/IF/SNK) (F019, F020).

Failure of these valves to shut represented a failure of both redundant safety divisions as PCIS valves F004 and F020 are associated with one logic division, and PCIS valves F003 and F019 are ascociated with the other logic division.

Following identification of the failure of these valves to close, each valve was given a manual close signal from the Control Roon reactor turbine gauge l board (EIIS/NA/CBD), at which time the F003 and F004 valves went shut. No change in position status was noted for the F019 and F020 valves. At >

approximately 0023 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months , the F020 was observed to be in the closed position.

No evolution could be identified which would have caused the F020 to close during this three minute time frame. Another manual close signal was given to ,

the F019 valve at approximately 0023 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months and again, no change in position i noted; however, the F019 valve was observed to be in the closed position at 0025 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months . Again, no evolution could be identified which would have closed the F019 valve following manual operation.

l Further review of the plant responsrv to this scram indicated that the remaining, plant safety systems operated per design. Five control rods (EIIS/AA/ ROD) were identified to be at the 02 position following the scram and they were fully inserted by 0030 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months using plant procedures.

Investination sn===ry Decreasing Condenser Vacuum j At approximately 2220 hours0.0257 days 0.617 hours 0.00367 weeks 8.4471e-4 months on 1/1/88, vacuum began to decrease following the l

removal of the 2B SJAE from service per procedure. Following an increase in i power and the placing of 2A SJAE in full load, vacuum began an improving <

trend. One hour later at 2345 hours0.0271 days 0.651 hours 0.00388 weeks 8.922725e-4 months , vacuum again began to decrease due to no apparent cause, as no plant evolutions had been undertaken within that hour I

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017 0 l5 0F ll0 which would have affected vacuum. Vacuum continued to decrease until the manual scram was initiated at 0017 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months , at which time vacuum recovered quickly, decreased for a short period of time, and then gradually increased until leveling out at 28 inches of Hg.

A review of the vacuum trend durin6 this event and previous plant vacuum operating history indicated that, although higher than normal air inleakage did contribute to the decrease in condenser vacuum, the high air inleakage '

flow rate would not alone cause the high SJAE discharge pressures observed during the event. It is unlikely that a larBe air inleakage source initiating at 2345 hours0.0271 days 0.651 hours 0.00388 weeks 8.922725e-4 months would have allowed the vacuum to recover and trend as noted following the scram.

As previously noted, air inleakage had been an identified problem prior to this event during power reductions. Investigations were initiated on 9/25/87 to identify and correct suspected air inleakage problems. The investigation consisted of a complete valve lineup and helium leak testing. These investigations had identified several air inleakage sources during the latter part of 1987 and were att11 in progress at the time of the event. Repairs had been made where system operation and safety would allow, with the remaining known air inleakage problems to be corrected during the outage. Additional air inleakage sources were still being sought at the time of this c. vent.

In addition to the known and suspected air inleakage, the decreasin6 trend in condenser vacuum is b.lieved to have been caused by SJAE back pressure. It is believed that excessive moisture in the downstream piping (r:IIS/SH/PSX) of the SJAEs was partially the cause for the decreasing vacuum. Indications of this probable cause included a high differential pressure alarm (possible cause; high moisture in the off gas effluent) received for the off gas system main stack filter (EIISW F/FLT) which cleared when the standby filter (EIISW F/FLT) was placed in service as well as a noted decrease in the off gas system recombiner (EIISWF/RCB) temperature. Moisture carryover into the discharge line (EIIS W F/PSX) may have occurred from the high level noted in the 2A SJAE intercondenser (EIIS/SH/COND), whi c was operating at a level of 82 inches (normal level is 55 inches).

During the Unit 2 1988 refueling / maintenance outage, extensive leak testing involving use of helium was performed on the main turbine and condenser and the interconnecting piping to the Condensate System (EIIS/SD) and the Main Steam Reheat System (EIIS/SB) as well as the main turbine structure to I determine the root cause of the incurred decreasing main condenser vacuum. On 4 February 2,1988, while the main condenser was intentionally flooded up, water was observed flowing from a discovered 3/4 inch hole in Miscellaneous Vents and Drains (MVD) System line 2.MVD 267 4 E 3 (EIIS/SM/PSX). This 4" line

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is the shell pocket drain line from the main turbine moisture separator reheater (MSR) to the main condenser, which collects the MSR shell drains from four 1 1/2" lines (EIIS/SM/PSX), each containing a 1/4" orifice strainer (EIIS/SM/PSX). When the main turbine is online., line pressure upstream of the strainers is that of the main turbine 7th stage extraction steam pressure, which varies with reactor power (80 psi 6 at $6% power to 170 psig at 100%

power. Likewise, line pressure downstream of the strainers varies with

reactor power, such that as power is decreased, the resulting reduced pressure in the line will cause greater main condenser air inleakage. The hole size was determined to have allowed an air inleakage of approximately 96 standard cubic feet per minute (scim) at a main condenser vacuum of 28 inches of mercury. Following discovery of the subject hole, the affected line was replaced. In addition to the hole in the MVD line, several other leaks were identified through helium testing. Repairs to these components were performed during the unit outage.

Valve Failures While performing a scram recovery in c.ccordance with plant procedures, it was determined that the group 2 PCIS valves 2 016.F003, F004, F019, and F020 failed to close on an automatic isolation signal (Iow level 1). Subsequent operator action caused the valves to close as previously stated. The following actions were performed in an effort to determine the cause of the PCIS valves' failure to close.

January 2,1988

1. Following the scram recovery, the four group 2 valves were successfully cycled during normal sump pumping operations with no problems noted with valve operation. The Unit 1 (U/1) valves were also stroked to verify operability.

2. A visual inspection was performed on the wiring (EIIS/JM/CBL1) and relays (EIIS/JM/RLY) associated with the group 2 isolation logic on Unit 2 (U/2).

3. A maintenance history search was initiated to develop the operatins history of the failed PCIS valves. This history review determined that the F003 valvc had experienced three failures and the F004 valve had experienced one failure since the solenoids (EIIS/IJ/*) were replaced in the spring of 1986 as part of environmental qualification modific.r~.ons.

The remaining two valves on U/2 and the four valves on U/1 did not have a failure histnry.

EIIS component description unavailable.

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4. Applicable sections of the logic system functional test procedures were performed, with no logic problems identified.

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- January 3, 1988

- 1. The F019 solenoid valve (ASCO) (E1IS/JM/PSV) was removed and disassembled. A minor oil film was identified on internal parts and minor debris was found in the solenoid valve. However , no cause for , or-evidence of, failure was identified.

2. The group 2 PCIS valve monthly functional test was performed on U/2. No problems were identified.

January 4, 1988

1. A special test procedure was performed to verify the group 2 PCIS valve system logic on U/1. No problems were noted.

January 6, 1"88

1. A spect test was performed to simulate the conditions present during the U/2 scram. The operation of the valves was observed locally and strip chart recording of electrical circuit operation was obtained. No problems were identified with valve operation.

2. The U/1 valves were shut (normally open), only to be opened for testing and sump pumping operations pending resolution of the failure of the U/2 valves.

January 7, 1988

1. Removed and disassembled the solenoid on the F003 valve with no problems idantified.

2. Performed an air cleanliness test of the air supply to the F003 valve and a test of the pressure regulator supplying the F003 valve. No problems were identified.

January 8, 1988

1. Removed the F020 valve solenoid for on-site vendor inspection and performed a pressure regulator test for its air supply. No problems were identified.

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1. Performed a visual inspection of the logic relays A71-K17 and A71 K18 1j (EIIS/JM/RLY), An are strike was identified on terminal 4 of the K18 relay and visual evidence indicates that contact-3-4 had been welded L closed as a result of the are strike, o

2. Initiated activities to remove one valve actuator (Miller air motor)- *

(EIIS/JM/84) for inspection. -

!l 3. The F020 valve solenoid, removed on 1/8/88, was inspectad on site by the l vendor with no. operability problems identified.

January 13, 1988

1. Removed and inspected the in line air filter (EIIS/JM/FLT) supplying the l F003 valve. No problems were idenH.fied (filter was very clean with approximately three years operating history). t January 14, 1988.

1. Replaced the A71-K17 and K18 relays, Ceneral Electric Part No.

CR120A06002AA, on U/2. s

.i January 19 and 20, 1988 '

1. Removed and inspected the valve actuator for the F020 valve. The i internals of the-actuator had a liberal coating of grease in accordance with vendor recommendations. 1

. Vendor inspection and analysis of the valves' solenoid valves did not identify evidence which explained.the failure of the valves to open. The inspection 3

'did reveal the presence of an oil bass film in the solenoid valves' internals; however analysis of the oil film deteimined it was oil used by the manufacturer for component assembly. Laboratory work at the. Carolina Power 6 i Light (CP&L) Company metallurgical and failure analysis ft.cility indicated the ;

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. incurred failure of the subject solenoids may have been the result of a higher '

than previously expected rate of degradation (oxidation) of the solenoids' 1 valve seat material, ethylene propylene diene monomer (EPDM), believed to occur when higher than anticipated ambient temperatures due to the solenoids' being normally energized, combine with the presence of copper in the brass solenoid valve bodies. In order to further identify the root cause(s) and required correction actions, the services of a contractor were retained.

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Brunswick Steam Electric Plant Unit 2 o p jo jo lo l 3j2 l4 8l8 0l0l 1 0 l7 0; 9 or 1 l0 inn m . wes maavim The results of the contractor's review were received during the first week of-June 1990. On July 15,1990, C.'6L Metallurgical Services Section issued Technical Report 13 23: Brunswick 2 Failure Analysis of Primary Containment, Isolation System (PCIS) Solenoid Valves. Based on the physical evidence, the report concluded that the solenoid valves failed to open due to adherence between the lower brass disc and EPDM seat with the most likely cause being copper-induced catalytic oxidation of the EPDM elastomer inserts. The report recommended that the Brutu. wick Plant use ASCO Model 206 832 solenoid valves . {

with Viton ciastomers for this application. It was also recommended that I I

additional characterization of EPDM materials he conducted through a utility funded organization such as the Electric Power Research Institute (EPRI) or the Boiling Water Reactor (BWR) Owners Group and that CP&L provide input to j the Nuclear Utility Group on Equipment Qualification (NUGEQ) allowing suggestions to be made to ASCO to pursue the investigation of the use_ of copper deactivators in their EPDM formulations if EPDM materials are continued to be used.

Corrective Actions f i

As.a result of this event, the on site Quality Assurance group performed a l surveillance activity (QASR 88-007) on the failure of the valves to close.

The solenoid valves for Unit 1 and Unit 2 PCIS valves C16-F003, F004, F019, and F020 have been replaced with the type solenoid recommended. In addition, ASCO solenoids with EPDM seats and normally energized coils'aro also being l replaced with solenoids having Viton seat material in accordance with !

Engineering Evaluation Request (EER) 88 076, Rev. 1. To date, twenty two i solenoids valves have been replaced and two remain. As an interim measure, the EER has directed that the two remaining solenoids be cycled on a weekly

^

basis to ensure their operability. The cycling is accomplished in accordance

. with Special Procedure 88-023.

In regard to actions recommended ir.volving EPRI, the BWR Owners Group, and NUGEQ CP&L corporate representatives for the associated' groups were contacted and are being provided a copy of Technical Report 13 23. They will review the report and initiate any additional actions.

Event Assessment This event was assessed to determine if the event would have been more severe under reasonable and credible alternative conditions as defined in NUREG 1022 (and supplements). This assessment determined that the first event (scram) would not have been more severe as this is an analyzed event in the safety analysis and the plant systems are designed for adequate mitigation. The second event (valve failures) would also not be more severe in that neither a reasonable nor a credible alternative conditidn could have provided a source term within the drywell. Without the source term, there is no increase in the quantity on material which would be released through these nonisolated penetrations.

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A review of our records indicates that the failure of the valves to close-is an isolated event for IIR reporting criteria (redundant system failure);

however, a condenser vacuum problem did initiate a scram in 1985 on U/1 and was reported in LER 1-85 008,

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05000324/LER-1988-001

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