ML20207T577

Insp Repts 50-348/87-02 & 50-364/87-02 on 870111-0210. Violation Noted:Failure to Adhere to Approved Procedures During Replacement & Cleaning of Diesel Generator Lube Oil Strainer Elements

ML20207T577
Person / Time
Site:	Farley
Issue date:	03/04/1987
From:	Brian Bonser, Bradford W, Dance H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207T563	List: ML20207T562 ML20207T571 ML20207T577
References
50-348-87-02, 50-348-87-2, 50-364-87-02, 50-364-87-2, NUDOCS 8703240147
Download: ML20207T577 (12)

See also: IR 05000348/1987002

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UNITED STATES -

'p# P K f 7pg'o NUCLEAR REGULATORY, COMMISSION

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g' j 101 M ARIETTA STREET.N.W.

• - '* ATLANTA. GEORGI A 30323

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Report Nos.: 50-348/87-02 and 50-364/87-02

-Licensee: Alabama Power Company -

'600 North 18th Street

Birmingham, AL 35291

Docket Nos.: 50-348 and 50-364 License Nos.: NPF-2 and NPF-8

Facility Name: Farley 1 and 2

Inspection Conducted: January 11 - February 10, 1987

Inspection at Fa'rley site near Dothan, Alabama

Inspec ' ors: - It' M 7/Y I7

Date Signed

y W . H. W Bradford

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Date Sig ed

Approved by:N

41. C. DanctA

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, Section Chief

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Odte Signed

Division of Reactor Projects

SUMMARY

Scope: This routine, onsite inspection included: monthly surveillance observa-

tion, monthly maintenance observation, operational safety verification, engi-

neered safety system inspection, licensee event reports, and onsite follow up

of events.

Results: One violation was identified in that approved procedures were not

adhered to during replacement and cleaning of diesel generator lube oil

strainer elements.

8703240147 870313

PDR ADOCK 05000348

G PDR

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REPORT DETAILS

1. Licensee Employees Contacted:

J. D. Woodard, General Plant Manager

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D. N. Morey,. Assistant General Plant Manager '

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W. D. Shipman, Assistant General Plant Manager 'o

R. D. Hill, Operations Manager

C. D. Nesbitt, Technical Manager

R. G. Berryhill, Systems Performance and Planning Manager

L. A. Ward, Maintenance Manager

L. W. Enfinger, Administrative Manager .i

B. Moore, Operations Supervisor

J. E. Odom,. Operations Unit Supervisor

B. W.'Vanland.ingham, Operations Unit Supervisor

T. H. Esteve, Planning Supervisor *-

J. B. Hudspeth, Document Control Supervisor

L. K. Jones, Material Supervisor

R. H. Marlow, Technical Supervisor

L. M. Stinson, Plant Modification Manager

J. K. Osterholtz, Supervisor, Safety Audit Engineeri.,g Review i

Other licensee employees contacted included technicians, operations

personnel, maintenance and I&C personnel, security force members, and

office personnel.

2. Exit Interview

The inspection - scope and findings were summarized during management

interviews throughout the report period and on February 12, 1987, with the

general plant manager and selected members of his staff. The inspection

findings were discussed in detail. One violation is discussed.in parag-

raph 6. The licensee did not identify as proprietary any material

reviewed by the inspector during this inspection.

3. Licensee Action on Previous Enforcement Matters (92702)

Not inspected.

4. Monthly Surveillance Observation (61726)

The inspectors observed and reviewed Technical Specification required

surveillance testing and verified that testing was performed in accordance

with adequate procedures; that test instrumentation was calibrated; that

limiting conditions were met; that test results met acceptance criteria

and were reviewed by personnel other than the individual directing the

test; that any deficiencies identified during the testing were properly

reviewed and resolved by appropriate management personnel; and that

personnel conducting the tests were qualified. The inspector witnessed /

reviewed portions of the following test activities:

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9- FNP-2-STP 61.0 .- Reactor Coolant Pump and.RHR Loop Operability

/i/erification. ',

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Notor Driven Auxiliary Feedwater Pump Check Valves '

~FNP-2-STP 22.12 -

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Flow Verification.

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3r FNP-2-STP 27.1 -

A. C. Source Verification. ,

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FNP-2-STP 27.3 -

Auxiliary and Service Water Building DC

Distribution; ~

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FNP-1-STP 227.5B - Radiation Monitor R24A Containment Purge and

Exhaust Isolation Functional Test. t

FNP-2-STP'35.0 -

Reactor Coolant System Pressure and Temperature

Limits Verification.

FNP-2-UOP 1.1 - Startup of Unit from Cold Shutdown to Hot Standby.

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TNP-2-UOP 1.1A 2 ~ Mode 4' Surveillance Check List. t

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No violatibns or deviations were identified.-

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'5. Monthly Maintenance Observation (62703)

Station ' maintenance activities of safety-related systems anc' componehts

were observed / reviewed to ascertain that they were conducted in accordance

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with approved propadures, regulatory guides, industry codes and standtirds,

4 and were in conformance with technical specifications. 1

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ys The following items were considered during the review: limiting condi-

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tions for operations aere met while components or systems were removed

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from service; approYsis were obtained prior to initiating the work;-

, j' activities were accomplished using approved procedures and were inspected

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as applicable; functional testing and/or calibrations were performed prior

to returning components or systems to service; quality control records a

were maintained; activities were accomplished by qualified personnel;

parts and materials were properly certified; radiological controls were

implemented; and fire prevention controls were implemented. Work requestss'

were reviewed to determine the status of outstanding jobs to assure that '

priority was assigned to safety-related equipment maintenance which may L'

affect system performance. The following maintenance activities were y

observed /rev,iewed: \

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. Unit 2 Excess Letdown Isolation Valve HV 8154 (MWR 145214)

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'28 Steam Generator Manway Cover Repair (MWR 151257[9 .

! . 5 \ 2C W esel Cenerator Maintenance Outage / Lube Oil Strainers and j

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Starting Air. PCN's

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. 10 and'1A Inverters

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Unit 2~ Plant Vent Stack Monitor Pump - R21 (MWR 145663).

. FNP-0'-MP-28.109 Siemens-Allis 4.16 KV Circuit Breaker PM's

No viola'tions:or deviations were identified.

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6. (Operational Safety Verification (71707)

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.The inspectors observed control room operations,_ reviewed applicable logs

and conducted discussions with control; room operators during-.the report:

period. The inspectors verified the operability . of selected emergency

rysta:as, reviewed tagout' records, and' vert fied proper return to service of

a1%3ed ' components. Tours of the auxiliary building, diesel building,-

turme . building and service water structure were conducted to observe

plact equipment conditions, including fluid leaks and excessive vibra-

- tions. De inspector verified compliance' with selected ' Limiting Conditions

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- _ . for. 0pem sons )(LCO) iand results of . selected surveillance tests. The

verifications' were accomplished by direct _ observation- of monitoring

instrumentation, . valve positions, switch positions,. accessible hydraulic

y snubbers; and review of completed logs, records, and chemistry results.

>The licensee's- compliance with LCO action statements were reviewed as

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events occurred.

The inspectors routinely attended meetings with certain licensee manage-

ment and observed various shift turnovers between shift supervisor, shift

foremen and licensed operators. -These meetings and discussions provided a

daily status of plant operations, maintenance, and testing activities in

progress, as well as discussions of significant problems.

'The inspector verified by observation and interviews with security force -

' members that measures taken to - assure the physical protection of the

. facility met current requirements. Ar'eas inspected included the organiza-

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. tion of the security force; the establishment and maintenance of gates,

doors, and. isolation zones; that access control and badging were proper;

and procedures were followed.

! ' On February 2,1987, while running IC diesel generator (DG) for normal

surveillance, the DG building System Operator (S0) noticed while taking

his ' logs that lube oil strainer differential pressure- (DP) was out of

specification high. The-IC #G strainer DP was 16 psid and the specifi-

' cation is-15 psid. The IC DG was shutdown to change lube oil strainers in

accordance with FNP-0-SOP 38.2 " Changing and Cleaning Lube Oil Strainer On

a D/G." During the process of cleaning the IC DG strainer elements the

~1icensee noticed that the strainer elements had numbers stamped on them

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. inappropriate for the 1C DG. S0P 38.2 step 4.3.3 specifies that 1C and 2C

DG's (Fairbanks - Morse) shall use strainer element number 200G (a 74

micron cylinder) and the 1-2A, 1B and 2B DG's (Colt - Pielstick) shall use

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strainer element number 2140 (a 40. micron cylinder). Upon further

. investigation- the licensee found incorrect strainer elements in other

DG's. The 1-2A DG had one 2140 strainer element and one 200G. strainer

element installed. 1B DG had two 200G strainer elements installed. 2B DG

had.two 2140 strainer elements installed - both correct; the 2C DG had two

2140 strainer elements and one 200G strainer element installed. A check

by the licensee of some recently completed SOP-38.2 procedures showed that

step 4.3.3 verification of proper strainer element part numbers had been

initialed as complete.

Strainer elements bearing incorrect part numbers were installed in DG lube

oil strainers and strainer element numbers were not verified subsequently

dur.ing cleaning. This is a violation for failure to follow an approved

procedure (348/364-87-02-01).

~The licensee has taken prompt corrective action to prevent reoccurrence of

these errors by changing the procedure to make it clearer .where the

numbers are located on the strainer elements and counseling operators to-

more carefully follow procedures. Also, the licensee has received corre-

spondence from Colt Industries indicating that use of the incorrect

strainers should not be an immediate safety concern on the 1-2A, 1B and 28

DG's but bearings should be checked during the next diesel maintenance

outage. There is no safety concern on the IC and 2C DG's since they use

the larger mesh strainer.

7. Licensee Event Reports (92700)

The following Licensee Event Reports (LERs) were reviewed ir potential

generic problems to determine trends, to determine wheths.- information

included in the report meets the NRC reporting requirements and to

consider whether the corrective action discussed in the report appears

appropriate. Licensee action, with respect to selected reports, were

reviewed to verify that the event had been reviewed and evaluated by the

licensee as required by the Technical' Specification; that corrective

action was taken by the licensee; and that safety limits, limiting safety

setting and LCOs were no exceeded. The inspector examined selected

incidents' reports, logs and records and interviewed selected personnel.

The following reports are considered closed:

Unit I

LER-83-09 - RHR pump inadvertently secured due to misinterpretation

of instructions.

LER-86-18 - Missed fire watches, personnel error.

LER-86-20 - Special Report: Opening of RCS Pressure Relief Valve.

LER-86-21 -

Special Report: Containment Equipment Hatch non-Functional

as Fire Barrier Longer Than Seven Days.

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LER-86-22:

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Mode change performed with RHR system automatic isolation

capability inoperable (personnel errors).

Unit II

LER-83-11 -

600 volt loud center declared inoperable due to procedural

inadequacy.

-The inspectors had no further questions.

8. -Follow up of_ Plant Events (93702)

On January 13,' 1987 at 7:13 a.m. Unit 'l was ramped down +.o 20%' power and

the generator taken off-line to repair a leaking turbine interface valve.

The valve was replaced and the unit returned to power operation at 2:45

p.m.

On January 22, 1987 the reactor tripped due to high flux rate as indicated

by nuclear instrumentation channels N41 and N42. The 1A inverter which

supplies power to N41 failed while N42 was in test for surveillance. This-

resulted in a two out of four coincidence for a high flux rate trip. The

testing on N42 was completed and N42 was returned to service. The backup

power supply for N41 was placed in service and the unit returned to power

operation on January 23, 1987.

Unit 2 returned to power operation on January 24, 2987 following a short

maintenance outage.

9. Engineered Safety Systems Inspection-(71710)

The inspector performed a system inspection of the diesel generators.

This inspection included the engine starting air system, engine jacket

cooling water system, service water system alignment in the diesel

generator building. diesel generator building fire protection and detec-

tion system, diesei generator building ventilation system, electrical

i switch gear alignment, annunciator response procedures, operating proce-

i dures, operator logs and housekeeping.

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The systems were assessed to be operable in accordance with the Technical

Specifications, appropriate drawings, procedures, and the Final Safety

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Analysis Report.

No violations or deviations were identified.

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10. Enforcement Conference Summary

On February 10, 1987, an Enforcement Conference was held in the offices.of

Region II to discuss the events surrounding the defeat of the- Residual

Heat Removal (RHR) system automatic isolation feature.

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Mr. L. A. Reyes opened the meeting by reviewing the NRC concerns.

Mr. R. P. Mcdonald gave his introductory statement and Mr. J. D. Woodard

gave a summary of events. Attachment 1 gives the agenda, system drawings

and applicable technical ~ specifications (TS) that were handed out by the

licensee.

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In the system description Mr. R. D. Hill provided a detailed description

of the RHR system and the Emergency Core Cooling System (ECCS). He

pointed out that the ECCS capability of the RHR subsystem was never

inoperable throughout this event. The valves that were improperly

jumpered are not utilized in the ECCS subsystem loop. The licensee took

exception to the wording used in inspection report 86-29 which referenced

the ECCS subsystem of being inoperable. They defined the problem in the

RHR system rather than ECCS. Additionally, he stressed that although the

automatic RHR isolation circuit was made inoperable, other system

interlocks which remained operable prevented the opening of the valves at

higher pressure. The licensee also expressed concern to the violation

refering to the lack of 10 CFR 50.59 review. The licensee did not agree

that due to a personnel error of not following procedures that inay fail

to perform a 50.59 review of a safety system modification.

Mr. J. D. Woodard gave his concluding remarks in that they believed the

root cause of the event was caused by personnel error. ' They admitted they

violated TS 3.0.3 and 3.0.4, however the significance of the violation

were minimum because the ECCS subsystem would have perform its intended

function.

Mr. R. P. Mcdonald gave his closing remarks that they take seriously all

issues relating to this event and that they will look at the maintenance

activities to see how they can be improved.

Mr. L. A. Reyes thanked the licensee for a well prepared presentation,

that it was very beneficial to all concern and that the matter will be

further reviewed by the Region.

Alabama Power Company Attendees

R. P. Mcdonald, Senior Vice President

J. D. Woodard, Plant Manager

L. A. Ward, Maintenance Manager

J. M. McGowan, Manager Safety Audit & Enigneering Review

R. D. Hill, Operations Manager

T. W. Cherry, I&C Supervisor

L. M. Stinson, Modifications Manager

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NRC Attendees

L. A. Reyes, Director, Division of Reactor Projects

G. R. lJenkins, . Director, EICS '

'H. C. Dance, Section Chief, DRP

E. Reeves,-Project Manager, NRR

F. Jape, Section Chief, DRS

' W. H. Bradford, Senior Resident Inspector, Farley

L. P. Modenos, Project Engineer, DRP-

' L. Trocine, Enforcement Specialist

B. Uryc, Enforcement Specialist-

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_. Attachment 1

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leC ENEGtCDENT CCMFERENCE

FEBRUARY 10, 1987

SUBJECT: RHR System Automatic Isolation and Interlock capability defeated

11/17/86 - 11/20/86

Alahnma Power Company Representatives

R. P. Mcdonald (Pat) Senior Vice President

J. D. Woudard (Jack) General Manager-Nuclear Plant

R. D. Hill (Richard) Manager-operations-Nuclear

L. A. Ward (Lewis) Manager-Maintenance-Nuclear

L. M. Stinson (Mike) Manager-Plant Modification &

Maintenance Support

T. W. Cherry (Ten) Plant Supervisor -

Instrumentation & Controls

AGEM R

I Introductions R. P. Mcdonald

II Summary of Events J. D. Woodard

III System Description / Technical

Specifications R. D. Hill

IV Sequence of Events L. M. Stinson

V Immediate Corrective Action L. A. Ward

VI Analysis of Event / L. A. Ward

Administrative Controls

VII Corrective Action J. D. Woodard

VIII NRC Concerns J. D. Woodard

IX Conclusion J. D. Woodard

X Closing Remarks R. P. Mcdonald

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RESIDUAL HEAT REMOVAL SYSTEM

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EMERCENCY CORE COOLING SYSTEMS

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3/4.5.2 ECCS SUBSYSTEMS - T

avg > 350*F

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LIMITING CONDITION FOR OPERATION

3. 5. 2

Two independent Emergency Core Cooling System (ECCS) subsystems shall be

OPERABLE with each subsystem comprised of:

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a. One OPERABLE centrifugal charging pump,

b. One OPERABLE residual heat removal heat exchanger,

c. One OPERABLE residual heat removal pump, and

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d.

An OPERABLE flow path capable of taking suction from the refueling

water storage tank on a safety injection signal and transferring

suction to the containment sump during the recirculation phase of

operatioa.

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APPLICABILITY: MODES 1, 2 and 3.

ACTION:

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a.

. With one ECCS subsystem inoperable, restore the inoperable subsystem

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'~' to 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 HOT STANDBY

N' 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 SHUTDOWN within the following

, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. In the event the ECCS is actuated and injects water into the Reactor

Coolant System, a Special Raport shall be prepared and submitted to

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the Commission pursuant to Specification 6.9.2 within 90 days

describing the circumstances of the actuation and tne total accumu-

lated actuation cycles to date. The current value of the usage

factor for each affected safety infection nozzle shall be provided

in this Special Report whenever its value exceeds 0.70.

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FARLEY-UNIT 1 3/4 5-3 AMENDMENT NO. 26

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EMERGENCY CORE COOLING SYSTEMS  %

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SURVEILLANCE REQUIREMENTS

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4,5.2 Each ECC5 subsystem shall be demonstrated CPERABLE:

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a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves

- are in the indicated positions with the disconnect device to the valve

operators locked open:

Valve Number . Valve Function Valve Position

s. 8884, 0886 Charging Pump Closed

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to RCS Hot Leg

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SD 2A, 8U 2B Charging Ptap Open*

b.

discharge isolation

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I c. 3889 RHR te RCS Hot Closed *

Leg Injection .

b. At least once Mr 31 days by verifying that each valve (manual,

power operatzd or automatic) in the flow path that is not locked,

. sealed, or othenvise secured ic position, is in its correct _)

position.

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l c. By a visual inspection which verifies that no loose debris (rags,

trash, clothing, etc.) is present in the containment which could be

transported to the containment sump and cause restriction of the

pump suctions during LOCA conditions. This visual inspection shall

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be performed: .

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1. For all accessible areas of the containment prict to

establishing CONTAINMENT INTEGRITY, and

. 2. Of the areas affected within containment at the completion of

ephcontainmententrywhenCONTAINMENTINTEGRITYis -

o'stablished. .

d. At least once per 18 months by: -

m5 1. VeNfying automatic isolation and interlock action of the

RHR system from the Reactor Coolant System when the Reactor

g:)[ , Coolant Systas pressure is between 700 psig and 750 psig.

2. A visual inspection of the containment sump and verifying that

the subsystem suction inlets are not restricted by debris and

that the sump components (trash racks, screens, inner cages)

are properly installed and show no evidence of structural

distress or corrosion. (3

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"Will be verified if charging pump 1A is declared inoperable.

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