ML20149G350

Safety Insp Repts 50-324/87-43 & 50-325/87-42 on 881201-31. Violations Noted.Major Areas Inspected:Maint Observation, Surveillance Observation,Operational Verification,Cold Weather Preparations & Onsite Followup of Events

ML20149G350
Person / Time
Site:	Brunswick
Issue date:	02/05/1988
From:	Fredrickson P, Garner L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20149G340	List: ML20149G337 ML20149G346 ML20149G350
References
50-324-87-43, 50-325-87-42, NUDOCS 8802180235
Download: ML20149G350 (12)

See also: IR 05000324/1987043

Text

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j d Mo UNITED STATES

p oq'o, NUCLEAR REGULATORY COMMISSION

J? ' ' ** REGION 11

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101 MAHlETTA STRE ET. N.W.

ATL ANT A, GEORGI A 30323

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Report No. 50-325/87-42 and 50-324/87-43

Licensee: Carolina Power and Light Company

P. O. Box 1551

Raleigh, NC 27602

Docket No. 50-325 and 50-324 License No. DPR-71 and DPR-62

Facility Name: Brunswick 1 and 2

Inspection Conducted: ecember 1 - 31, 1987

Inspectors- b' ,

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W. H. Rtland Date Signed

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Date Signed

Approved By: G <

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P. E. Fredtickson, Section Chief

~he / ' ' 7/C/k

Dhte Signed

Division of Reactor Projects

SUMMARY

Scope: This routine safety inspection by the resident inspector involved the

areas of followup on previous enforcement matters, maintenance observation,

surveillance observation, operational safety verification, cold weather

preparations, and onsite followup of events.

Results: In the areas inspected, one violation was identified: failure to

deactivate primary containment system isolation valves.

8802180235 880200 4

PDR ADOCK 0500

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DETAILS

1. Persons Contacted

Licensee Employees

• E. Bishop, Manager - Operations

T. Cantebury, Mechanical Maintenance Supervisor (Unit 1)

• G. Cheatham, Manager - Environmental & Radiation Control

R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

• C. Dietz, General Manager - Brunswick Nuclear Project

• W. Dorman, Supervisor - QA

• R. Eckstein, Manager - Technical Support

• K. Enzor, Director - Regulatory Compliance

• R. Groover, Manager - Project Construction

• W. Hatcher, Supervisor - Security

A. Hegler, Superintendent - Operations

R. Helme, Director - Onsite Nuclear Safety - BSEP

J. Holder, Manager - Outages

• P. Howe, Vice President - Brunswick Nuclear Project

L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)

R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

• J. Moyer, Manager - Training

• J. O'Sullivan, Manager - Maintenance

B. Parks, Engineering Supervisor

R. Poulk, Senior NRC Regulatory Specialist

J. Smith, Manager - Administrative Support

R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

D. Warren, Acting Engineering Supervisor

B. Wilson, Engineering Supervisor

Other licensee employees contacted included construction craftsmen,

engineers, technicians, operators, office personnel, and security force

members.

• Attended the exit interview

2. Exit Interview (30703)

The inspection scope and findings were summarized on January 5, 1987, with

those persons indicated in paragraph 1. The inspectors described the

areas inspected and discussed in detail the inspection findings listed

below. Dissenting comments were not received from the licensee. The

licensee did not identify any information supplied to the inspector as

proprietary.

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Note: Acronyms and abbreviations used in the report are listed in para-

graph 9.

Item Description / Reference Paragraph

325/87-42-01 & VIOLATION - Failure to Deactivate Primary

324/87-43-01 Containment Isolation Valves (paragraph 3).

325/87-42-02 & *URI - Wrong Unit Event Involving RHR Pump

Breakers

324/87-43-02 (paragraph 6.a).

324/87-43-03 URI - Reactor Coolant System Leakage in MSIV Pit

(paragraph 8.d).

325/87-42-04 URI - Dial Type Thermometar in SLC Tank

(paragraph 4.b).

325/87-42-05 URI - Vital Area Access To Service Water Valve

324/87-43-05 Pits (paragraph 8.b).

324/87-43-06 URI - RHR SW Gasket Leak (paragraph 8.c).

325/87-42-07 IFI - Lonegren SLC Relief Valve Plug Installed

with Incorrect Drain Plug (paragraph 4.a).

325/87-42-08 & IFI - Submission of TS Amendment Request for RWCU

324/87-43-08 Isolation Response Time (paragraph 6.b).

I 325/87-42-09 & IFI - Diesel Generator Building Supply Fan "A"

Failure (paragraph 8.a).

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324/87-43-09

325/87-42-10 IFI - HPCI F001 Motor Failure (paragraph 4.c).

3. Followup on Previous Enforcement Matters (92702)

(CLOSED) Unresolved Item (325/87-39-04 and 324/87-40-04), Deactivation of

Primary Containment Isolation System Valves. Discussions with NRR on this

issue confirmed that "deactivate" as used in TS means that automatic

valves must be rendered incapable of operating, thus preventing both

inadvertent and spurious operation. The licensee failed to properly

deactivate valves on three occasions:

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• An Unresolved Item is a matter aoout which more information is required to

determine whether it is acceptable or may involve a violation or deviation.

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a. The gas sample return to suppression pool inboard valve, 2-RXS-SV-

4188, had dual indication during performance of PT-4.1.1, Reactor

Building Ventilation Exhaust Monitoring System Functional Test,

Revision 31. The licensee performed the test on October 6, 1987, and

declared the above valve inoperable at 4:30 a.m. the same day. To

comply with TS 3.6.1 1, the licensee, under clearance 2-1044, red-

tagged closed 2-RXS-SV-4189 the same day. The valve was still

capable of being opened by the control switch. The clearance was

closed on December 25, 1988.

b. The reactor water sample line isolation valve, 1-B32-F019, had failed

indication during performance of PT-3.1.22, Reactor Coolant Recir-

culation System Valve Operability Test, Revision 10. The licensee

had performed the test on September 22, 1987, and declared the above

valve inoperable at 10:44 p.m. the same day. To comply with TS 3.6.3, the licensee, under clearance 1-1809, red-tagged closed

1-B32-F020, the outboard valve, the same day. Fuses were not pulled

and the valve was still capabla of being opened by the control

switch. Corrective action was taken on December 9, 1987.

c. The drywell head inboard purge exhaust valve, 1-CAC-V49, had no

position indication during routine surveillance te sti ng . The

licensee had performed the test on June 16, 1987, and declared the

above valve inoperable at 5:00 a.m. the same day. To comply with TS 3.6.3, the licensee, under clearance 1-1618, red-tagged closed

1-CAC-V50, the outboard valve, the same day. Circuit breakers were

not opened and the valve was still capable of being opened by the

control switch. Corrective action was taken on December 9, 1987.

The above failures to deactivate the containment isolation valves is a

Violation: Failure to Deactivate Primary Containment Isolation Valves

(325/87-42-01 and 324/87-43-01).

One violation and no deviations were identified.

4. Maintenance Observation (62703)

The inspectors observed maintenance activities, interviewed personnel, and

reviewed records to verify that work was conducted in accordance with

approved procedures, Technical Specifications, and applicable industry

codes and standards. The inspectors also verified that: redundant

components were operable; administrative controls were followed; tagouts

were adequate; personnel were qualified; correct replacement parts were

used; radiological controls were proper; fire protection was adequate;

quality cor. trol hold points were adequate and observed; adequate post-

maintenance testing was performed; and independent verification require-

ments were implemented. The inspectors independently verified that

selected equipment was properly returned to service.

Outstanding work .equests were reviewed to ensure that the licensee gave

priority to safety-related maintenance.

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The inspectors observed / reviewed portions of the selected maintenance

activities throughout the month. The matters below require followup,

a. SLC Relief Valve Drain Plug

A drain plug on the 1A SLC pump discharge relief valve, 1-C41-F029A,

blew out while the system was idle on December 8, 1987, at 10:15 a.m.

The resulting leak from the 5/8 inch nole was due to the head from

the SLC tank. The tank was isolated at 10 : 19 a .m. , stopping the

leak. The plug was replaced by 1:16 p.m. the same day.

The licensee found that the drain plug was not made of stainless i

steel as required, but appeared to be a carbon steel pipe plug. The

plug was not listed on the mechanical drawing for the valve. The

relief valve was made by Lonegren with a stainless steel body and

internals. The valve had been installed since April 1987. The pipe

plug was found with severely corroded threads, possibly from galvanic

corrosion. A valve in stores also had a similar plug. The licensee

verified that the other 3 valves in the plant had the correct plugs.

The Harris E&E Center will examine the plugs to determine the exact

material. The inspector will follow the licensee's resolution of the

issue, particularly any reportability determination. This is an

Inspector Followup Item: Lonegren SLC Relief Valve Plug Installed

with Incorrect Drain Plug (325/87-42-07).

b. Thermometer in SLC Tank

A five inch dial type thermometer was removed from the Unit 1 SLC

tank at 1:15 p.m. on December 23, 1987. The licensee determined that

f the system operability would not be affected in that the thermometer

was underneath the sparger at the tank bottom and could not have

migrated to the tank outlet line. The inspector will follow the

licensee's review of this event, including root cause determination.

This matter remains Unreselved pending the inspector's followup of

the licensee's root cause analysis: Dial Type Thermometer in SLC

Tank (325/87-42-04).

c. HPCI F001 Motor Failure

On December 31, 1987, at 1:00 a.m., the HPCI steam admission valve,

1-E41-F001, failed to open during performance of surveillance test

PT-9.2. The licensee found the motor armature grounded, the shunt

field open and the motor internals blackened. In,pection of the

motor control center, in situ testing of the torque switch and

actuation of the valve by application of a torque wrench to the hand

wheel drive has revealed no mechanical cause for the failure.

Portions of the initial investigation conducted under work request

87-BMTII, OSPP-BKR004 and MP-57, were observed by the inspector. The

inspector plans to continue to follow the licensee's efforts in this

area, specifically regarding root cause failure analysis. This is an

Inspector Followup Item: HPCI F001 Motor Failure (325/87-42-10).

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No violations or deviations were identified.

5. Surveillance Observation (61726)

The inspectors observed surveillance testing required by Technical

Specifications. Through observation, interviews, and record review, the

inspectors verified that: tests conformed to Technical Specification

requirements; administrative controls were followed; personnel were

qualified; instrumentation was calibrated; and data was accurate and

complete. The inspectors independently verified selected test results and

proper return to service of equipment.

The inspectors witnessed / reviewed portions of the following test activi-

ties:

2MST-RPS11W Main Steamline High Radiation Channel Functional Test.

2MST-RWCU21M RWCU High Differential Flow Trip Unit Channel Calibration.

PT-01.11 Core Performance Parameter Check.

PT-14.1 Control Rod Operability Check (Unit 1). .

No violations or deviations were identified.

6. Operational Safety Verification (71707)

The inspectors verified that Unit 1 and Unit 2 were operated in compliance

with Technical Specifications and other regulatory requirements by direct

observations of activities, facility tours, discussions with personnel,

reviewing of records and independent verification of safety system status.

The inspectors verified that control room manning requirements of 10 CFR

50.54 and the Technical Specifications were met. Control operator, shift

supervisor, clearance, STA, daily and standing iastructions, and jumper /

bypass logs were reviewed to obtain information concerning operating

trends and out of service safety systems to ensure that there were no

conflicts with Technical Specifications Limiting Conditions for Opera-

tions. Direct observations were conducted of control room panels,

instrumentation and recorder traces important to safety to verify oper-

ability and that operating parameters were within Technical Specification

limits. The inspectors observed shift turnovers to verify that continuity

of system status was maintained. The inspectors verified the status of

selected control room annunciators.

Operability of a selected Engineered Safety Feature division was verified

weekly by insuring that: each accessible valve in the flow path was in

its correct position; each power supply and breaker was closed for compon-

ents that must activate upon initiation signal; the RHR subsystem cross-

tie valve for each unit was closed with the power removed from the valve

operator; there was no leakage of major components; there was proper

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lubrication and cooling water available; and a condition did not exist ,

which might prevent fulfillment of the system's functional requirements. '

Instrumentation essential to system actuation or performance was verified

operable by observing on-scale indication and proper instrument valve

lineup, if accessible.

The inspectors verified that the licensee's health physics policies /

procedures were followed. This included observation of HP practices and a

review of area surveys, radiation work permits, posting, and instrument

calibration.

The inspectors verified that: the cecurity organization was properly

manned and security personnel were capable of performing their assigned

functions; persons and packages were checked prior to entry into the

protected area; vehicles were properly authorized, searched and escorted

within the PA; persons within the PA displayed photo identification

badges; perscnnel in vital areas were authorized; effective coinpensatory

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measures were employed when required; and security's response to threats

or alarms was adequate.

The inspectors also observed plant housekeeping controls, verified

position of certain containment isolation valves, checked clearances at

random, and verified the operability of onsite and offsite emergency power

sources,

a. Wrong RHR Pump Breaker Cleared

On December 21, 1987, at about 4:00 a.m. an operator racked out the

wrong RHR pump breaker while hanging a clearance. The operator had

intended to rack out the 2A RHR pump breaker on 4160 V emergency bus

E-3. Instead, the operator racked out the 1A RHR pump breaker on the

same bus. The operator discovered his mistake when he returned to >

the control room and found the 2A pump's control room light still

energized. The licensee immediately declared the Unit 1 Division I

RHR system inoperable and racked in the 1A pump breaker, cancelling

the LCO. Both units were in Operational Condition One at the time of

the event. The inspector's final resolution of this issue will be

made pending the licensee's completion of their OER. This matter

remains Unresolved: Wrong Unit Event Involving RHR Pump Breakers

(325/87-42-02 and 324/87-43-02).

b. Technical Specification Discrepancy

The licensee identified, in June 1985, that the isolation system

i instrumentation response time for the RWCV high flow isolation was

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inaccurate. Table 3.3.2-3, item 3.a, listed the response time as 5

13 seconds. However, this does not include the 45 second timers in

the circuit. The instrumentation compares the inlet and outlet f",ows

to determine possible leakage. The licensee stated that the total

time for the instrument response (45 + 13 = 58 seconds) is below the

time used in the GE analysis for a break in the RWCV line.

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Based on discussions with Region 'II, the li;ensee was asked to submit

the TS amendment that was planned over two years ago. During the

exit interview, the licensee committed to submit the amendment

request by March 31, 1987. This matter is an Inspector Followup

Item: Submission of TS Amendment Request for RWCU Isolation Response

Time (325/87-42-08 and 324/87-43-08).

No violations or deviations were identified.

7. Cold Weather Preparations (71714)

The inspector verified that the licensee had implemented 01-43, Freeze

Protection and Cold Weather Bill, Rev. 1, on December 29, 1987. The

inspector verified that the freeze protection circuit lights were

energized for the RCIC/HPCI condensate storage tank low level switches.

The inspector noted that the thermometer used to measure ambient tempera-

ture in the service water building was missing. The shift foreman

directed an auxiliary operator to replace the thermometer.

No violations or deviations were identified.

8. Onsite Followup of Events (93702) .

a. Diesel Generator Building Supply Fan Failure

A diesel generator building supply fan failed catastrophically on

December 13, 1987. The fan was made by Joy Manufacturing Company and

has an airfoil blade design. The fan is one of four fans that

supplies ventilation to the four site emergency DGs. Only three fans

are needed, per FSAR section 9.4.7.2, to maintain adequate ventila-

tion during worst case conditions. The eight blades of the 66 inch

fan broke up into hand sized pieces. The pieces were sent to CP&Ls

laboratories. Preliminary findings indicate that six of eight blades

had undergone significant fatigue failure. Microscopic cracks were ,

found at the blade base on the opposite side of the fatigue failure.

! The licensee has about 60 Joy fans onsite of the same design. This

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is the first failure of this type. The OG fan had had its first

hand-held vibration readings taken on December 1, 1987, with no

readings above the manufacturer's shutdown limit. No vibration

signature analysis had been done on this fan or any other diesel

supply or exhaust fan. The licensee also found a bent nickel (five

cents) in the fan housing.

The licensee is repairing the damaged fan. Once the fan is repaired,

the licensee plans to perform dye penetrant testing on the other

diesel supply fans as well as continue their metallurgical analysis.

Cracks were also found in the air flow straightening vanes adjacent

to the hub in DG fan "A". The licensee made weld repairs on two

vanes and removed the crack tip by drilling on the remaining cracks.

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A monthly inspection program has been established for the "A" fan,

which includes a full spectrum analysis. An inspection of the

remaining fans is also scheduled for completion by March 18, 1988, as

part of an action item in EER 87-0549. In the exit interview, the

licensee committed to the inspector to complete the inspections by

the above date and to provide a date for completion of the metallur-

gical analysis ir one week. This is an Inspector Followup Item:

Diesel Generator Building Supply Fan "A" Failure (325/87-42-09 and

324/87-43-09).

b. Vital Area Access

The inspector reviewed a security event with the licensee's security

supervisor. The event involved an individual previously approved for

unescorted access to the Central Alarm Station who entered the CAS

while not on the current access list. Further discussions with the

security supervisor raised questions concerning access controls for

other vital areas. The licensee put administrative controls in place

on December 28, 1987, that resolved the inspector's immediate

concerns. Region II security inspectors will resolve this matter

during subsequent inspections. This matter is Unresolved: Vital

Area Access. (325/87-42-05 and 324/87-43-05).

c. Unit 2 RHR SW Leak

During operation of the Division I Residual Heat Removal Service

Water system, a gasket failed on 2-E11-F014A, the RHR HX SW inlet

valve. The failure occurred at 9:30 a.m. on December 24, 1987, with

the unit at 70% power coasting down to refueling. The licensee was

making preparations to run RCIC by starting the Division I RHR SW and

RHR system in the torus cooling mode. The RHR pump had been started

shortly before the gasket failed. The control operator imnediately

stopped the RHR pump when he received the report of the leak from the

Auxiliary Operator. The A0 then determined the exact source of the

leak and the control operator secured the RHR SW pump. An upstream

valve was torqued shut to completely stop all leakage.

The RHR 2A and 2C pumps were sprayed with salt water and about 2,000

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gallons (six inches) of water remained in the Division I RHR -17 foot

area. The leak had occurred on the 20 foot level, near the top of

the HX. The licensee declared Division I of LPCI, RHR SW, and

Suppression Pool Cooling inoperable. The licensee's reco/ery actions

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included leak repair, inspection and testing of wetted equipment, and

removal of the water by pumping to 55 gallon drums. The drums were

then emptied into the salt water release tank and released through

the service water effluent line, All affected systems were declared

operable by 7:54 p.m. on December 27, 1987.

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RHR Division II was unaffected. The senior resident inspector was

outside the RHR HX room when the gasket failed. The inspector

concluded that the licensee's actions during the event were appro-

priatt. The licensee plans to followup the event with an OER. Final

resolution of this matter awaits inspector review of the OER. This

matter remains Unretolved: RHR SW Gasket Leak (324/87-43-06).

d. Unit 2 MSIV Pit Leak

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The licensee discovered, on December 13, 1987, a five to 10 GPM leak

coming from between the concrete encased torus roof and the floor of

the MSIV pit tunnel. The licensee found the leak rate by recording

the pump run time in the reactor building south core spray sump.

Reactor building humidity increased as shown by condensation on HVAC

ducts and the RBCCW heat exchangers. No radioactive airborne problem

was noted. The licensee reported that' chemistry sampl's of the water

indicated, at that time, that the leak was from main steam.

The licensee commenced valve isolations on December 15 to locate the

leak. Per SP-87-100, each main steamline was isolated, one at a

time, with no change in leakage, which had increased to 15 GPM by

then. The licensee isolated the MSL drains also with no change ir,

leakage. On December 17, 1987, the licensee electrically bacKseated

the outboard feedwater system stop-check isolation valves, 2-B21-

F032A and B, stopping the leak.

The inspector questioned how primary containment integrity was being

maintained with the valve operators electrically backseated. The

licensee modified the valve logic, initiated procedure controls, and

computed potential leakage during valve ycling under those controls

to show that containment integrity was maintained.

The licensee temporarily modified the valve logic, providing long

! term containment isolation capability. The F032 check valves will

! automatically shut on a feedwater line break outside containment.

For long term isolation, the operator may be required to shut the

valves using the operator, maintaining the penetrations shut.

With the current packing leak, the licensee would be required to shut

the check valves firmly by driving the disk into the seat. Once

again the valves would have to be backseated. To electrically

backseat the valves, the open limit switches must be bypassed. The

licensee, in EER 87-0550, December 18, 1987, bypassed the open limit

switches, to permit electrically backseating the F032 valves from the

control room if the reactor building was inaccessible during an

accident. Standing instructions were provided to the operators on

how to operate these valves under those conditions.

The licensee concluded that the packing leakage that would occur

during valve cycling was acceptable. The supporting analysis was

performed in EER 87-0551, Revision 1, December 23, 1987. The deter-

mination of acceptability was based on:

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o The . oquivalent air leakage rate + previous ILRT results +

backseated leakage rate was less than the allowable leakage rate

for containment for one hour.

o Standing instructions to operators for valve operations.

o Continued monitoring of south core spray pump room sump with an

established maximum leakage rate.

The inspector concluded that the licensee's above actions were

acceptable.

The inspector further questioned compliance with TS 3.4.3.2, Reactor

Coolant System Operational Leakage. Part of the TS limits RCS

leakage to five GPM UNIDENTIFIED LEAKAGE averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

period. Before the leakage was identified on December 17, the ,

licensee operated several days with a leak that was unidentified and

may have been coming from the RCS. FSAR page 3.1.2-27 lists the

boundary of the RCS as the P032 valves. However, the surveillance

requirements for this TS only list drywell related equipment to

monitor for leakage. The leakage was shown to be from the pack'ng of

the F032 valves, part of the reactor coolant system boundary.

Therefore, TS 3.4.3.2 should apply. The licensee had not considered

whether the TS had applied prior to tha inspector's question. This

matter remains Unresolved pending further Region II/NRR review:

Reactor Coolant System Leakage in MSIV Pit (324/87-43-03).

No violations or deviations were identified.

9. List of Aobreviations for Unit 1 and 2

A0 Auxiliary Operator

BSEP Brunswick Steam Electric Plant

CAS Central Alarm Station

DG Diesel Generator

EER Engineering Evaluation Report

ERFIS Emergency Response Facility Information System

ESF Engineered Safety Feature

F Degrees Fahrenheit

FSAR Final Safety Analysis Report

GE General Electric Company

GPM Gallons Der Minute

HP Health Physics

HPCI High Pressure Coolant Injection

HVAC Heating, Ventilating, Air Conditioning System

HX Heat Exchanger

I&C Instrumentation and Control

IE NRC Inspection and Enforcement

IFI Inspector Followup Item

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ILRT Integrated Leak Rate Test

LCO Limiting Condition for Operation

LER Licensee Event Report

LPCI Low Pressure Coolant Injection

MP Maintenance Procedure

MSIV Main Steam Isolation Valve

MSL Main Steamline

NFC Nutt.ar Regulatory Commission

N,;R Office of Nuclear Reactor Regulation

OER Operating Experience Report

01 Operating Instruction

PA. Protected Area

PNSC Plant Nuclear Safety Committee

PT Periodic Test

QA Quality Assurance

QC Quality Control

RBCCW Reactor Building Closed Cooling Water

RCIC Resctor Core Isolation Cooling

RCS Reactor Coolant System

RHR Residual Heat Removal

RWCU Reactor Water Cleanup

SLC Standby Liquid Control .

SP Special Procedure

STA Shift Technical Advisor

SW Service Water

TS Technical Specification

URI Unresolved Item

V Volt

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