ML20236G452

Insp Repts 50-313/87-30 & 50-368/86-30 on 870901-30. Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint,Surveillance,Followup on Previously Identified Items & Followup on Allegation

ML20236G452
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/19/1987
From:	Craig Harbuck, Jaudon J, Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20236G444	List: ML20236G441 ML20236G447 ML20236G452
References
50-313-87-30, 50-368-87-30, NUDOCS 8711030015
Download: ML20236G452 (20)

See also: IR 05000313/1987030

Text

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APPENDIX B ' >3 . .-

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? Inspection-Report: :50-313/87-30i '

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Licensee:P.i0.1 Arkansas, Box 551'.' '

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Inspections ummary

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Inspection Conducted: September 1-30, 1987'(Report 50-313/87-30)

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Areas Inspected: Routine,' unannounced inspection including operational safety

verification, maintenance, surveillance, followup on previously identified

items, followup on an allegation, and 10 CFR Part 21 reports.

Results: 'Within the six areas inspected, one apparent violation was identified

(inadequate preventive maintenance program for lubricating pump couplings,

paragraph 5).

Inspection Conducted September 1-30, 1987 (Report 50-368/87-30)

Areas Inspected: Routine, unannounced inspection of operational safety

. verification, maintenance, surveillance, followup on previously. identified

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items, followup on an allegation, Technical Specification 4.8.1.1.2.c.12, and

10 CFR Part 21 reports. i

Results: Within the seven areas inspected, one apparent violation was

identified (inadequate preventive maintenance program for lubricating pump

couplings, paragraph 5).

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DETAILS

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1. Persons Contacted

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• J. Levine, Executive. Director, Site' Nuclear Operations

B. Baker, 0perations Manager

E. Bickel, Health Physics Superintendent  ;

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J. Bruni, Shift Maintenance Supervisor

B. Cameron, Bechtel Lead Engineer

A.. Cox, Unit 1 Operations Sup'erintendent 4

! E. Corliss, Instrumentation and Controls Supervisor

R. Douet, Qua'lity Assurance Auditor

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M. Durst, Project Engineering Superintendent

R. Dyer, Planning and Scheduling Coordinator  ;

- *E. Ewing, General Manager, Technical Support l

B. Garrison, Operations Technical _ Support 1

D. Graham, Quality Control Engineering Supervisor i

• H. Green, Quality Assurance Superintendent ,

L. Gulick, Unit 2 Operations Superintendent  !

C. Halbert, Engineering' Supervisor

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-J.~ Hale Bechtel Engineer

A. Hatley, Mechanical Maintenance Supervisor i

H. Hollis, Security Superintendent  !

• D. Howard, Special Projects Manager -;

• L. Humphrey, General Manager, Nuclear Quality

G. Kendrick, Instrumentation and Controls Superintendent j

J. Lamb, Safety and Fire Prevention Coordinator

• R. Lane, Engineering Manager

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• D. Lomax, Plant Licensing Supervisor ,

A. McGregor, Engineering Services Supervisor q

D. McKenney, Engineer <

• J. McWilliams, Maintenance Manager j

C. May, Acting Mechanical Maintenance Supervisor ,

• P. Michalk, Licensing Engineer '

D. Payne, Maintenance Coordinator _

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• V. Pettus, Mechanical Maintenance Superintendent

S. Quennoz, General Manager, Plant Operations

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P. Rehm, Mechanical Maintenance Engineering' Technician

P. Rogers, Special Projects Coordinator j

C. Shively, Plant Engineering Superintendent 1

• M. Smith, Reactor Engineering Supervisor l

C. Taylor,. Unit 2 Operations Technical Support Supervisor

• J. Taylor-Brown, Quality Control Superintendent 1

L. Taylor, Special Projects Coordinator -  !

R. Wewers, Work Control Center Manager

M. White, Engineering Technician

B. Williams, Engineering Supervisor

T. Windham, Bechtel Engineer l

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G., Wrightam, Instrumentation and Controls Supervisor

C.. Zimmerman, Unit 1 Operations Technical Support Supervisor

'*Present at exit interview.

The NRC inspectors also contacted other plant personnel including <

operators, technicians, and administrative personnel.

2. Followup on Previously Identified Items (Units 1 and 2)

(Closed) Unresolved Item.368/8633-01: FLilure to Meet the Original f

Acceptance Criteria for the Reference Control Element Assembly (CEA) Bank

Reactivity Wocth - The licensee resolved this. deficiency, as discussed in

NRC. Inspection Report 50-313;368/86-33, by changing the acceptance criteria .le

such that it encompassed the. measured. bank worth, with a temporary change

to Procedure 2302.03, " Determination of CEA Worths by Exchange." The NRC

inspector referred the supporting licensee safety evaluation to the NRC

Office of Nuclear Reactor Regulation (NRR). for' review. NRR concluded that

additional justification should be provided by the licensee.

The following points formed the basis of the licensee's safety evaluation:

a. All of the test CEA bank reactivity worths were determined by l

measurement to be within their acceptance criteria. '

b. Apparently a systematic deviation had affected either the measurement i

or prediction since all but one bank were more reactive than predicted.

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c. One input to the process of determining the predicted bank worth was

the bias applied to the computer calculated worth. (This bias will be

further discussed below). For the Cycle 6 predicted worth

determination, the bias had been increased over the value used in

previous cycles. This had the effect of lowering the'value of the

predicted worths more than in'past cycles,

d. The Final Safety Analysis Report (FSAR) applied a 10 percent

uncertainty to the assumed worth of the most reactive CEA for both .

dropped and ejected CEA accident analysis. Thus, the actual worst  !

case reactivity worth for each accident was still bounded by the FSAR l

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From these four points, the licensee concluded that the increased worth of

the reference bank would neither increase the probability nor'the

consequences of any the accidents analyzed in the FSAR; additionally the  ;

other safety criteria of 10 CFR 50.59 would not be violated. However, for l

verification of this conclusion, the safety eval ation' stipulated that  !

further review be done following the completion of the 30 percent and

50 percent power physics testing.

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The results of the' .. staff'.s. review of the ' safety l evaluation were as follows:

The staff' found that Item a. above'was .a' valid point; especia11y 'since it . J'

was oneiof the three acceptance criteria for the -Combustion Engineering rod l

swap method'of measuring _CEA bank worths discussed in Topical-  :{

Report CEN-319-A. Further,3 1t was noted that the second of the three.  !

acceptance 1 criteria could have been mentioned. since it was.also met. This l

. criteria was the total, worth of all the CEAs. -The one criteria ~ not met was

the measured reactivity worth'of the reference bank. j

Items b and c were~consilered to be related.; Apparently the systematic. q

deviation was the bias applied toithe computer' generated worth. This had

affected the predicted: values. The Cycle.6 calculated CEA bank worths.had ,

been determined by an NRC approved reactor core' simulator' computer . , . 1

code.(ROCS). These:results were then' modified by the application of.a bias ' i

expressed as a perce.ntage of calculated worth. Generally, the bias'.is

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determined by statistically comparing the computer code generated results

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to historical measurement data. Application of the bias is meant to G

improve the accuracy of the predicted CEA bank worths. However, for

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Cycle 6, the fuel management core-loading pattern was changed. Based on.

l limited measurement data 'obtained from other Combustion Engineering reactor i

cores utilizing similar fuel management strategies and enrichments, the  !

bias had been increased for Cycle 6. A~ recent reevaluation of this bias,

using additional measurement data, now indicates that a lower value.of bias

is more appropriate (4 percent instead of 7 percent). Had this .. lower bias

value been.used for the Cycle 6 predictions, the reference bank worth

acceptance criteria ( 10 percent)'woul_d have been met.

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The staff found that Item (d) contributed nothing to the basis of the

l safety evaluation because one cannot infer the'effect'of an increase in a

CEA bank reactivity worth on the accident analysis .by examining the' .. .

analysis uncertainties in the assumed worth of an individual CEA in the l

FSAR accident analysis. -j

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l Additional support for the safety analysis was provided from examination of' l

i the core power distribution measurements at 30 percent, 50 percent,'and

L 100 percent power. These measurements indicated that the portions of the

core adjacent to the reference-bank CEAs were slightly:more reactive than q

predicted by Combustion Engineering. This.would have the effect.of .;

increasing the reactivity worth of the reference bank. ll

The NRC staff pointed out two additional. items which support the conclusion

of the safety evaluation.

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Previous staff reviews of a number of rod swap methodologies: indicate

that the error which. occurred on the reference bank (2 percent beyond-  !

the allowed plus/minus percent band) had little' effect on the overall i

CEA bank measurement results.

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Examination of the measured critical positions' with the . estimated ' ,

critical' positions-(and the critical boron concentration for the i

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reference bank) for all the CEA banks showed good agreement except for  ;

one test bank. However, this test bank had a measured reactivity  ;

' worth equal to the predicted value. '

- Based on the above, the NRC staff concluded that the higher than predicted

reactivity worth of the reference bank _ posed no safety problem; that the

licensee's initial safety evaluation, although not comprehensive in its 1

basis, had reached an acceptable conclusion; and that no violation of NRC l

requirements had occurred. l

However, the NRC staff pointed out that the Combustion-Engineering Topical i

Report, CEN-319-A " Control Rod Group Exchange Technique," approved by the j

NRC, did not directly address the action to be taken should the reference

bank measurement fall- outside the acceptance criteria. The staff suggested .-

that a better way to have resolved the problem in the Cycle 6. testing would - -)

have been to define the bank with the highest predicted reactivity worth as  ;

the new reference bank and measure its' worth by boron dilution. The. 1

' remaining test banks could then have been measured using this new reference

bank by the rod swap method. This resolution method is allowed by the

Topical Report.

The NRC inspector discussed this point with the licensee. The licensee

committed to revise Procedure 2303.03 to include the contingency method

noted above should the reference' bank measurement fall outside the

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acceptance criteria in future cycle startup physics testing. The NRC

inspector also suggested that NRR be consulted should activities in the i

future result in the desire to revise acceptance criteria which have been i

specifically approved by the NRC staff (such as the plus/minus'10' percent

measured to predicted value of the reference bank _ stated in-Topical

Report CEN-319-A).

Based on the above, this item is closed. _]

(Closed) Unresolved Item 313/8718-02: Basis for Process Monitor

Calibration Procedure - The NRC inspector reviewed the initial test

procedure for the Unit 1 radiation monitoring system which was performed by

the contractor, LFE Corporation, in 1971. The licensee's current

calibration procedure is essentially equivalent to that procedure.

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Further, the licensee appeared to have an adequate understanding of the

technical aspects of the calibration process. The NRC inspector concluded

that the licensee has an adequate basis for the Unit 1 process monitor

l calibration procedure. This item is closed.

l (0 pen) Unresolved Item 368/8525-01: Process Radiation Monitor Calibration

- The principle concern of this item was the need to improve the

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measurement of the plateau region of the gaseous monitor Geiger-Muller

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detectors. Data from the recent calibration (June-August 1987) of a number

1 of these monitors was reviewed to ascertain whether any improvements had'

been made. The procedure used 2304.27, " Process Radiation Monitoring

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System Calibration," Revision 14, was essentially unchanged from the j

previous revision used in the 1985 calibrations. Three problems were noted

by the NRC inspector.

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. It appeared that in several cases the plateau region extended beyond

the upper limit.of data collection,1000 Vdc, set b

This item has been corrected as the new procedure (y the procedure..

discussedbelow)

requires measurements up to 1200, Vdc.

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. What apparently was interpreted'as.the plateau ~ region of .. .. .

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Monitor 2RITS-8846, contained a-peak. This had not been evaluated by'~ j

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the licensee.

. Monitor 2RITS-8845 high voltage had b'een adjusted to a value below the

voltage range of the plateau during the calibration voltage adjustment.

Monitors 2RITS-8846 and -8845, which monitor the penetration room emergency

ventilation ' exhaust, are not req'uired by the Technical Specifications' (TS).

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These two items were.promptly' addressed by the licensee. Using a new l

procedure (2304.06, " Gaseous Process Radiation Monitoring System

Calibration," Revision 1, approved September 17,1987) the monitors were

recalibrates as part of Job Order 738245. The NRC inspector reviewed the..

data. The high voltage had previously been set around 850 to 900 Vdc. i

This time both monitors were set at 1147 Vdc. At the~ time.of the review,

the licensee had reached no conclusion about the implications of this large

adjustment in voltage. The licensee has plans to redo all the gaseous

process monitors under the same job order in the near future due to .

problems with the old procedure. Further review to. include'the results of

these additional recalibrations is needed before this item can be' resolved.

Therefore, this item remains open.

3. Operational Safety Verification (Units 1 and 2)

The NRC inspectors observed control room operations, reviewed a]plicable

logs, and conducted discussions with control room operators. T1e NRC

inspectors. verified the operability of selected emergency systems, reviewed

tagout records, and ensured that maintenance requests had been initiated-

for equipment in need of maintenance. The NRC inspectors made spot checks

to verify that the physical security plan was being implemented. The NRC.

inspectors verified implementation of-radiation protection controls 'during

observation of plant activities.

The NRC; inspectors toured accessible. areas of.the units to observe plant

equipment conditions, including potential fire' hazards, fluid leaks, and

excessive. vibration. The NRC inspectors also observed plant' housekeeping

and' cleanliness conditions during the tours.

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'The NRC inspectors walked down the accessible-portions of.the Unit 1-

service water system to verify operability. The walkdown was conducted

using Procedure 1104.29, Attachment "A", . Revision .26, and Drawing M-210,

Revision 47. No significant problems were~noted.-

During routine tours, the NRC inspector made the following observations:

. Stream Trap 2F-340 Isolation Valve 2MS-97-2 had a severe ' packing leak.

The licensee was informed. . Job Request 986631 was issued to stop' l

the leak. This steam trap is in the line which drains the line just

after Valve 2CV-1000-1, the "A" steam generator supply isolation.to

Emergency Feedwater Pump 2P7A's . turbine driver, 2K3.

. The cable cover to 2 LIT-4908 (Boric Acid Makeup Tank'2T6B level..

switch) was' torn. : This was repaired promptly by. the licensee.

While touring the P36C makeup pump room the NRC inspector observed a

seismic pipe support that did not conform to its design configuration. .The- .

hanger, HBD-20-SW-2, mounted -vertically from the ceiling, supports the -

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service water 3-inch diameter supply line to Room Cooler VUC-7C. OneLof

the concrete expansion anchors had apparently broken free and had dropped  ;

almost completely out of its hole. After being informed, the licensee ' '

issued a Report of Abnormal Conditions (RAC 1-87-167) and Job

Request 786933 to repair the hanger. The NRC inspector told the licensee '

that this particular noncompliance should be considered as another example j

of previous similar violations. The NRC inspector noted that a system 'I

walkdown program committed to by the licensee as long-term preventive

corrective action for such violations, was just beginning.' This program is  ;

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designed to include identification of configuration control and degradation '

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problems of safety-related piping systems including the seismic supports..

The adequacy of this program will be evaluated during a future inspection.

The NRC inspector accompanied a health physics supervisor on a' routine tour

of. the Unit 1 auxiliary building. It was noted.that the contaminated

controlled area around the Duratek filter equipment was in need of

housekeeping attention. This problem was soon corrected. The NRC

inspector noted no other significant problems, and considered the tour to

have been conducted in a thorough manner.

These reviews and observations were conducted to verify that facility-

operations were in conformance with the requirements established under TSs,

10 CFR, and administrative procedures.

No violations or deviations were identified.

4. Monthly Surv'eillance Observation (Units'1 and 2)

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The NRC inspector observed the TS required surveillance testing on Unit 2

Charging Pump 2P36A (Procedure 2104.02, Supplement I) and verified that

testing was performed in accordance with adequate procedures, test .

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instrumentation was calibrated, limiting conditions for operation were met,

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removal and restoration of the affected components were' accomplished, test

results conformed with TSs and procedure requirements, test results were

reviewed by personnel other than the.. individual-directing the. test, and any ,

deficiencies identified during the . testing,were properly reviewed and ,l

resolved by appropriate management personnel.

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The'NRC inspector also witnessed portions of the following' test ' activities:- ,

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. Hydrogensamplertest(Procedure.1104.31, Supplement .1)' j

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. . Test of motor-driven emergency feedwater pump after coupling  !

inspection (Procedure 1106.06, Supplement I) l

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. Station battery pilot cell' tests (Procedure 1307.16)

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.- Monthly test-of emergency diesel generator (Procedure 1104.36,

l Supplements)

. Test of emergency diesel generator to prove operability following i

failure of the other emergency diesel generator (Procedure 1104.36, .;

Supplement II)

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. Margin to saturation instrument' channel calibration. (Procedure 1304.84,

Job Order 740691) j

.. Test of emergency diesel generator to prove operability following

maintenance (Procedure 1104.36, Supplement:II)

. Reactor building cooling coil service water flow test

(Procedure 1104.33, SupplementVI)  ;

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l . ' Monthly test of high pressure injection pump (Procedure 2104.39, .

SupplementIII) I

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. Monthly test of low pressure injection pump (Procedure 2104.40,

Supplements)

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. Calibration check of reactor building pressure instrument supplying

Channel "C" of the reactor protection system (Procedure 1304.43)

. Monthly test of emergency diesel generator (Procedure 2104.36,

SupplementII)

No violations or deviations were identified.

5. Monthly Maintenance Observation (Units l'and 2) j

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Station maintenance activities of safety-related systems and components'

listed below were observed to ascertain that they.were conducted in

accordance with. approved procedures, Regulatory Guides, and industry codes

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or standards; and in 'conformance with TSs.

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The following. items were considered during this review: . the limiting'

conditions -for operation were met while components or systems were. removed

from service, approvals were obtained prior to initiating the work,  :

activities were accomplished using approved procedures and wereLinspected l

as' applicable, functional testing and/or' calibrations were performed prior -

to returning components or systems to service, quality control records were  :

maintained, activities were. accomplished by qualified personnel, parts and '

materials used were properly certified, radiological controls were

implemented, and fire prevention controls were implemented.

Work requests were reviewed to determine status'of outstanding jobs and to  !

ensure that priority is assigned to safety-related equipment maintenance l

which may affect system performance. 1

The following maintenance activities were observed: ,

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. Replace Auxiliary Relay 152-408/X (Job Order 739521) ]

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. Replace motor to gear coupling on P36C (Job Order 739489, i

Procedure 1402.010) j

. Replace leaking discharge flange gasket on high pressure safety l

injection pump (Job Order 737405)

. Replace starting air pressure regulators on emergency diesel generator l

(JobOrder 735623)  ;

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. Replace air start solenoid valves on emergency diesel generator (Job j

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Orders 727428 and 735454)

. Diesel fire pump quarterly surveillance inspection (Procedure 1306.27,

Job Order 738311)

. Inspection of pump end coupling of Charging Pump 2P36A (Job

Order 740070)

. Inspection of high pressureinjectionpumpcoupling(JobOrder 740070,

Procedure 2402.36)

On September 1,1987, the NRC inspector observed the disassembly of the

motor to gear coupling on the "C" makeup pump on Unit 1. It was suspected

that this coupling had failed since operators had observed the motor

running and the pump not rotating after the pump lost discharge pressure.

The coupling disassembly and inspection were performed under Job

Order 739489. The coupling gears were found to be badly worn, and only a

small amount of hardened grease was found in the coupling. Several' days

later, after obtaining a new coupling and revising Procedure 1402.010 to

provide maintenance guidance, the coupling was replaced.. Under the same

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job order, the NRC inspector observed the l inspection of'the gear to pump

coupling on the "C" makeup pump. This coupling and-its grease were found

to be in good condition.

The NRC ins Sector reviewed the technical manual for the makeup pumps, Byron

Jackson Tec1nical Manual G404550, and discussed with licensee personnel how

the motor to gear coupling limited the motor end. float. It was determined

that buttons or spacer discs are not needed to limit end float due to the

clearances established when the gear and motor were mounted on the base

plate.

Step 7.18.2 of Procedure 1402.010 includes instructions to install. spacer

plates and buttons in the motor to gear coupling. Since these are not

needed and not used, licensee personnel stated that this section of the

procedure would be revised. In addition, several other sections of this

procedure will be revised to incorporate the temporary changes which were

made prior to installation of the new motor to-gear coupling. This item

will renmin open pending revision of Proceduro 1402.010 and NRC inspector

review of the revised procedure. (313/8730-02)

During the followup of the P36C motor coupling failure event, the NRC

inspectors determined that the licensee's preventive maintenance program

for safety-related pump couplings-was deficient, in that the lubrication

schedules were not well defined and apparently not followed. This

conclusion was based on the licensee being unable to find lubrication

records of several pump couplings subsequent to the date indicated below:

. High Pressure Injection (HPI) Pump P36A motor No record

coupling

. HPI Pump P36C pump coupling July 1979

. HPI Pump P36C motor coupling 1982

. Low Pressure Injection (LPI) Pump P34A May 1982

. LPI Pump P34B No record

. Reactor Building Spray (RBS) Pump P35A- April 1979

. RBS Pump P35B No record

. Emergency Feedwater Pump 2P7B No record

Inadequate lubrication was apparently the primary contributor to the P36C

motor coupling failure.

Lubrication schedules for safety-related equipment are specified in

Regulatory Guide 1.33, Appendix "A", which is comitted to by TS 6.8.1 of

both units. . Failure to establish and implement adequate lubrication

preventive maintenance schedules is an apparent violation. (313;368/8730-01)

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The NRC inspectors noted that the licensee pursued an aggressive program i

to identify ar.d correct possible generic problems indicated by.the P36C

coupling failure. This program included: ,

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a. Inspection of all high' speed lubricated cou

pumps (and also on nonsafety-related pumps)plings

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on safety-related

This effort was j

completed for' safety-related pumps on Septembar'14, 1987. No 1

additional lubrication problems were noted. 1

b. Determination of the. status of the safety related preventive ..

maintenance program.' This effort was scheduled to be completed by

October 9,1587. Overdue PMs identified will be scheduled as j

appropriate. -

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The NRC inspectors will continue to follow the licensee's implementation J '

of this program. The licensee was informed that the generic implications'

of the P36C coupling failure, already being pursued as just noted, and the

corrective actions taken as a result, should be addressed in the' response

to the violation noted above.

6. Followup of Allegation 4-86-A-042 (Units 1 and 2)

The following concern obtained from the review of statements provided to 1

the NRC, was addressed during this inspection. 1

An alleger stated that-at unspecified times and locations unqualified

workers used sharp stainless steel hooks to remove foam from electrical

conduits without deenergizing the cables in the conduits. Discussions

with site personnel indicated that poor practice in'this area had been a

problem in the past. The NRC. inspector reviewed-a memo dated July 18,

1984, which stated.a policy that no sharp metal tools would be used for

penetrating and removing foam. The NRC inspector reviewed the file on a

personnel injury accident which occurred on October' 24,1984. In this

case, an engineer was injured when using a nylon probe with a. metal hook

attached to one end to check the depth of.a foam dam installed in a

conduit under a motor control center. When the probe was removed, the

metal contacted the feeder bus in the motor control center causing a short

resulting in a fireball and severe personnel injury. Following this

incident, a policy of' requiring an approved work plan for penetration

sealing work in energized electrical equipment was adopted. This policy 4

required consideration of physical barriers and electrical-isolation.

The NRC inspector reviewed Procedure 4033.06,~'! Installation, Repair and/or

Alteration of Fire Barrier Penetration Seals." This procedure requires

that qualified personnel perform penetration seal work and ' includes

requirements that only nonconducting instruments shall be used when

working with electrical equipment and that no sharp or metal tools shall'

be used in removing old sealants from around cables.

The NRC' inspector concluded that the allegation was probably correct and

that the poor practice resulted in a personnel injury in October 1984.

___

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.i

13

Since that time, controls over penetration sealing have been improved and.

they appeared to be adequate at the time of this inspection.

This. allegation remains open pending the review of the other technical

concerns identified during the review of the statements provided to the

NRC.

No violations or deviations were identified.  ;

7. Interpretation of TS Surveillance Requirement 4.8.1.1.2.c.12 (Unit 2)

TS 4.8.1.1.2.c;12 states, "Each diesel generator shall be demonstrated

OPERABLE: At least once per 18 months during shutdown by: Verifying that 1

with the diesel generator operating in a test mode (connected to its bus),  ;

a simulated safety injection signal overrides the test mode by' i

(1) returning the diesel generator to standby' operation and

(2) automatically energizes the emergency loads with offsite power."

During the last performance of this surveillance, the NRC inspector had

commented that the licensee's procedure did not appear to meet the intent

l of this TS in that no loads were automatically started as described in (2)

i above; but rather the actuation of relays in the breaker closing circuits

for the associated emergency loads were observed and timed. At that time,

the licensee revised their procedure to require the actual starting of

some, but not all, of the loads possible. This.was considered an

.

'

acceptable alternative at the time by the NRC inspector.

During this inspection period, preparation of the surveillance test

proc 2 dure to be used in the upcoming 1988 refueling outage was in

progress, and the licensee requested additional technical guidance on the

correct interpretation of this TS, from the NRC inspector.

! -A conference call between the licensee, the NRR project manager, an NRR

l technical staff representative, and the NRC inspector. was held on

September 22, 1987, to discuss this TS. The licensee was. told by NRR that

'

the proper . interpretation of Item (2) of the TS was to automatically

l energize all the emergency loads that were possible with offsite power.

l It was suggested that a change of this interpretation could be. pursued by

i

means of a TS amendment proposal. <

Persons participating in the conference call were:

NRR AP&L E Site

l Jim Knight Larry Taylor Craig Harbuck

! George Dick Don Lomax

No violations or deviations were identified.

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'8. 'Part 21' Reports (Units l'and 2)-. ,

,

During review of 10 CFR Part 21 reports submitted ^by manufacturers,;

suppliers, and reactor licensees, NRC personnel identified certain. reports. -!

which.could be applicableLat ANO. The NRC inspector provided . copies of'

these: reports,to the licensee. The. reports provided1are: listed in the

' ' '

attached ..tabl e. It is. expected that the licensee will review these-

~

reports.to' determine whether they are applicable to ' equipment'at ANO. g

- , J

Part 21 Reports Provided to '.icensee- 1

Number Originator' Date Subject ]

-86-002 Georgia' Power Co. 09/05/86' -Pipe support' tolerance.

?andLinsta11ation  :

-procedures

i

86-003 Indiana & Michigan 09/18/86 . Defective emergency.

Electric Co.- head lever. supplied .for a' ux .

t

feedi. pump :.

!86-009 Georgia Power Co. 07/31/86 ESpring failure-Valcor

l solenoid valves *

.86-013 Foxboro Co. 10/07/86- ' Advisory on handling ,

Foxboro N-E111and N-E13

l transmitters-

,87-002  ; Virginia Electric 11/12/86 t Potential defect in.new svc

- and Power Co. ' water spray support. system

-

>

87-003 Foxboro Co. 06/04/86. End of Life susceptibility

of e-line &.H-line. i

'

instruments

i

87-004 Indiana & Michigan 12/20/85- Weld electrodes with

Electric'Co. incomplete flux coating

87-005 Florida Power 07/08/86 Tip damage on. anti-reverse

& Light Co. ,

rotation device pins.on RCP.87-006 Portland General 07/25/86 Stationary sleeve on MSIV .

-

Electric Co. thrust bearings interference

87-007 -PROMATEC, Inc. 02/17/86' Defective conduit seals-

87-011 .Gensral Electric Co.. 11/17/86 HFA relays could experience.

incorrect. operations:

87-016 Limitorque Corp. 12/19/86' ' Damaged insulation lon? ._  ;

Limitorque valve operator DC c

motor

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87-019' Vermont Yankee '11/10/86 Design defect in'Limitorque ,

Nuclear Power Corp. valve operators pre-1975

87-020 Automatic Sprinkler 12/01/86- Automatic fire ' sprinkler ,

Corp. system valve failure j

87-025 GA Technologies, Inc. 02/23/87 Low insulation' resistance.off

coax cable for HRR monitors87-028 Niagara Mohawk '01/26/87 Improper seating'of Agastat ,.

GP series relays ,

(87-029 Toledo Edison 02/03/87 Inadequate instructions to

maintain torque switch

'

.

' balance

87-030 Niagara Mohawk 02/02/87 ' Improper electrical

-

duct seal design

87-031 Automatic Valve 12/19/86 Houghto 620 lubricant

l

Corp. attacks, & degrades aluminum

'

valves

l

87-035 Foxboro Co. 02/17/87 Foxboro spec 200 C/V cards 'j

affected by high moisture '87-036 Sacramento Municipal 02/10/87 Limitorque warped limit

Utility District switch rotors'-

87-038 Morrison-Knudsen 01/13/87 .EDG control ll relay failed to l

Co., Inc. drop out when deenergized- .)

.I

87-044 Arizona Nuclear 03/02/87 Replacement fuel injection. .j

Power Project tube nuts not per SAE J521b j

87-046 Isomedix 03/30/87 EQ qualification

questionable.

1

9. Exit Interview

I

The NRC inspectors met with Mr. J. M. Levine, Executive Director, ANO Site

Operations, and other members of the AP&L staff at the end of the-

inspection. At this meeting, the NRC inspectors summarized the scope of

the inspection and the findings.

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