ML20203F429

Insp Repts 50-313/86-21 & 50-368/86-22 on 860601-30. Violation Noted:Failure to Follow post-maint Test Procedure

ML20203F429
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/16/1986
From:	Craig Harbuck, Hunter D, Johnson W, Vissing G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20203F391	List: ML20203F387 ML20203F407 ML20203F429
References
50-313-86-21, 50-368-86-22, IEB-85-003, IEB-85-3, NUDOCS 8607310065
Download: ML20203F429 (11)

See also: IR 05000313/1986021

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION

. REGION IV

NRC Inspection Report: 50-313/86-21 Licenses: DPR-51

- - 50-368/86-22 NPF-6

Dockets: 50-313

50-368

, Licensee: Arkansas Power and Light Company (AP&L)

- 'P. 0.'. Box 551

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- - Little Rock,~ Arkansas 72203

Facility 'N'ame: Arkansas Nucl' ear One (ANO), Units 1 and 2

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Inspection Nt: dNOSite,'Russellville, Arkansas

Inspection Conducted: June 1-30, 1986

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Inspectors: I 7

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e V. D. Johnsor6 SeMaf Resident Reactor

J N Inspector

(pars. 2, 3, 4, 5, 6, 7)

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pC' C. Harbuck, R6siqWfit Reactor Inspector

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g g . S. Vissing,# Pref 5ct Manager

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Approved:

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D. R. Hunter, Chief, Reactor Projects Date

Section B, Reactor Projects Branch

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PDR ADOCK 0500

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Inspection Summary.

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' Inspectio'n Conducted June 1-30, 1986 (Report 50-313/86-21)

, - TAreasInspected: Routine,' unannounced inspection including operational safety

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verification, maintenance, surveillance, followup on previously identified

items, followup on Licensee Event Reports, followup on IE Bulletin 85-03,

review of transient reports, and 10 CFR 50.59 evaluations.

Results: Within the eight areas inspected, one violation was identified

(failure to document safety evaluations as required by 10 CFR 50.59,

paragraph 9).

Inspection Summary

Inspection Conducted June 1-30, 1986 (Report 50-368/86-22)

Areas Inspected: Routine, unannounced inspection including operational safety

verification, maintenance, surveillance, followup on Licensee Event Reports,

followup on IE Bulletin 85-03, 10 CFR 50.59 evaluations, and refueling

activities.

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

(failure to follow post-maintenance test procedure, paragraph 6).

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'ETAILS

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1. Persons Contacted

J. Levine, Director of Site Nuclear Operations

R. Ashcraft, Electrical Maintenance Supervisor

• B. Baker, Operations Manager

R. Bennett, Consultant, Wyle Laboratories

J. Brown, Quality Assurance Supervisor

• P. Campbell, Licensing Engineer

• M. Cooper, Quality Assurance Inspector

A. Cox, Operations Technical Support Supervisor

• E. Ewing, General Manager Technical Support

M. Goodson, Civil Engineer

L. Gulick, Unit 2 Operations Superintendent

H. Hollis, Security Coordinator

D. Horton, Quality Assurance Manager

• D. Howard, Special Projects Manager

L. Humphrey, General Manager, Nuclear Quality

D. Johnson, Licensing Engineer

H. Jones, Field Construction Manager

P. Kearney, Project Engineer

• R. Lane, Engineering Manager

• D. Lomax, Licensing Supervisor

B. Lovett, Electrical Maintenance Engineer

A. McGregor, Engineering Services Supervisor

• J. McWilliams, Maintenance Manager

V. Pettus, Mechanical Maintenance Superintendent

D. Provencher, Quality Engineering Supervisor

E. Rice, Electrical Maintenance Supervisor

P. Rogers, Plant Licensing Engineer

C. Shively, Plant Engineering Superintendent

R. Simmons, Planning and Scheduling Supervisor

C. Taylor, Operations Technical Support Supervisor

R. Taylor, Consultant, Wyle Laboratories

B. Terwillicer, Operations Assessment Supervisor

D. Wagner, IIealth Physics Supervisor

• R. Wewers, llork Control Center Manager

C. Zimmerman, Operations Technical Support

• Present at exit interview.

The inspectors also contacted other plant personnel, including operators,

technicians, and administrative personnel.

2. Followup on Previously Identified Items (Unit 1)

(Closed) Deviation 313/8523-01: Failure to meet a commitment relative to

implementation of 88W Safe-End Task Force recomendations.

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The licensee's response to this item included a clarification of the

commitments made in their letter to the NRC dated April 22, 1985. The

licensee provided this clarification and commitment revision to the NRC

Office of Nuclear Reactor Regulation in a letter dated March 10, 1986.

3. Licensee Event Report (LER) Followup (Units 1 and 2)

Through direct observation, discussions with licensee personnel, and

review of records, the following event reports were reviewed to determine

that reportability requirements were fulfilled, immediate corrective

action was accomplished, and corrective action to prevent recurrence has

been accomplished in accordance with Technical Specifications.

Unit 1

.86-004, Reactor Trip

Unit 2

85-024, Fire Barrier Seal Degradation

86-001, Reactor Trip

Unit 1 LER 86-004 reported a reactor trip on high reactor coolant system

pressure due to a turbine control system malfunction. The root cause was

believed to be a lightning strike which disturbed the turbine control

system through the plant grounding system. The NRC inspector reviewed

Report of Abnormal Conditions (RAC) 1-86-070 and the job order used to

troubleshoot, repair, and test the turbine control system.

Unit 2 LER 85-024 reported degradation of the fire barrier seal between

the auxiliary building floors and the reactor building wall. This barrier

was sealed and added to the fire barrier surveillance program. ,

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Unit 2 LER 86-001 reported a reactor trip due to a spurious closure of a

main steam isolation valve. The NRC inspector reviewed RAC 2-86-018 and

the job orders used to replace the relays suspected of causing the valve

closure. (JO 708120 and 708228)

No violations or deviations were identified.

4. Operational Safety Verification (Units 1 and 2)

The NRC inspectors observed control room operations, reviewed applicable

logs, and conducted discussions with control room operators. The

inspectors verified the operability of selected emergency systems,

reviewed tagout records, verified proper return to service of affected

components, and ensured that maintenance requests had been initiated for

equipment in need of maintenance. The inspectors made spot checks to

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verify tl.at the physical security plan was being impleirented in accordance

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with the station security plan. The inspectors verified implementation of l

radiation protection controls during observation of plant activities. l

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The NRC inspectors toured accessible areas of the units, including the

Unit 2 containment building, to observe plant equipment conditions,

including potential fire hazards, fluid leaks, and excessive vibration.

The inspectors also observed plant housekeeping and cleanliness conditions

during the tours.

During a plant tour, the NRC inspector noted that pipe hanger 20CB-3-H9

on emergency feedwater piping, downstream of 2FE-0717 and upstream of

2CV-1075-1, was carrying no load. The hanger drawing for this hanger

indicated a dead weight load of 197 pounds downward. This condition was

identified to the licensee, and the hanger was adjusted.

While touring the Unit 2 reactor building, the NRC inspector found

deficiencies in several seisnic class I pipe supports. These deficiencies

were subsequently documented by the licensee under two RACs (Report of

Abnormal Conditions) as follows:

(1) RAC 2-86-121 'A' steam generator

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. loose baseplate for pipe hanger supporting line 2DBB-1000-3/4"

containing 2SGS-21A, a level transmitter lower tap isolation

valve.

. loose anchor bolts on pipe support 2DBB-7-H5 which supports

line 2DBB-7-4" the steam generator blowdown line.

(2) RAC 2-86-122 'B' reactor coolant pump

. loose anchor bolts on the two pipe supports adjacent to each

side of 2RCP-3B, the controlled bleedoff manual isolation valve.

The licensee issued Job Requests 15488 and 15489 to correct these

deficiencies and planned to perform an engineering evaluation of their

effect on operability of the supports. These deficiencies will remain as

dn open item pending review of the licensee repair of the supports and of

the completed operability evaluation. (368/8622-01)

The NRC inspector observed that two of the Unit 1 service water pump

discharge pipes were corroded, as were the discharge strainer body for P4C

and the piping flange bolts for P4A. A licensee representative stated

that a job order had been issued to remove the corrosion and paint the

affected areas.

The NRC inspectors walked down the accessible portions of the Unit 1 and

Unit 2 emergency boration flowpaths. The walkdown was performed using

Procedures 1104.03, 2104.03, 1202.01, and 2202.01 and Drawings M-231,

M-233, and 11-2231. No system lineup discrepancies were identified but

minor errors were noted on Drawing M-233. These were identified to the

licensee for correction.

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, These reviews and observations were conducted to verify that selected

facility operations were in conformance with the requirements established

under Technical Specifications, 10 CFR, and administrative procedures.

No violations or deviations were identified.

5. . Monthly Surveillance Observation (Units 1 and 2)

The NRC inspector observed that the Technical Specification required

surveillance testing on the systems listed below and verified that testing i

-was performed in accordance with adequate procedures, test instrumentation '

was calibrated, limiting conditions for operation were met, removal and

restoration of the affected components were accomplished, test results

conformed with Technical Specifications 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 resolved by appropriate management personnel.

The inspector witnessed portions of the following test activities:

. Monthly channel test of reactor protection system channel A

(Procedure 1304.37) (JO 712931)

. Engineered safeguards actuation system analog channel number 1 test

(Procedure 1304.49) (JO 713104)

. Escape lock barrel test (JO 712934)

. Hydrogen purge system test (Procedure 1104.33, Supplement II)

j . Service water flow test (Procedure 2311.02)

. Test of control room emergency ventilation unit VSF-9

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(Procedure 1104.34, Supplement 1A)

l . Monthly channel test of reactor protection system channel C

, (Procedure 1304.39) (JO 713858)

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. Station battery 2012 service discharge test (Procedure 2403.26)

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(JJ 708158)

18-month operability test of the 'B' diesel generator

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(Procedure 2104.36, Supplement 4)

, . Overspeed test of the 'B' diesel generator (Procedure 2104.36,

Supplement 7)

. EFW pump P7A monthly test (Procedure 1106.06, Supplement 2)

No violations or deviations were identified.

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6. Monthly Maintenance Observation (Units 1 and 2)

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Station maintenance activities of safety-related systems and components

listed below were observed to ascertain that they were conducted in

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accordance with approved procedures, Regulatory Guides, and industry codes

or standards; and in conformance with Technical Specifications.

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 were inspected

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

l 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

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which may affect system performance.

The following maintenance activities were observed:

. Service water pump preventive maintenance (Procedure 2402.34)

, (JO 711288)

. . Repair Target Rock solenoid valve position indication

(Procedure 1403.02) (JO 713486)

. Diesel generator fuel solenoid valve inspection (JO 0233)

. . 18-month inspection of diesel generator (Procedure 2306.05)

(JO 707885)

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. C high pressure injection pump motor preventive maintenance

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(JO 525514)

. Condensate storage tank piping (Design Change Package 82-2086B)

(JO 703886)

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. Condensate. storage tank foundation and pipe chase (DCP 82-2086)

! . Uncouple motor from C high pressure injection pump (JO 711764)

. Anchor bolt replacement (DCP 84-2043) (JO 711454)

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. Battery charger 2034 load test following maintenance

. (Procedure 2403.53) (JO 714504)

- The NRC inspector found that the lineup specified in Procedure 2403.53 for

testing battery charger 2D34 was not being used during the

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post-maintenance test. The procedure lineup requires that 2D34 be loaded

by bus 2001 or 2002 and that a testing unit be used to obtain a load of

200 amps on the battery charger. During the test observed on June 27,

1986, battery charger 2034 was not connected to either battery bus, and

the testing unit was supplying the total load. Breaker 72-0211, which was

specified to be closed by the procedure, was tagged open; and the green

power leads from 2D34 to 2002, which are not addressed in the procedure,

were lifted. No procedure change to permit this lineup had been obtained.

This is an apparent violation. (368/8622-02)

7. Followup on IE Bulletin (IEB) 85-03 (Units 1 and 2)

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IEB 85-0_3 Gas' issued on November 15, 1985, and is entitled,

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" Motor-0perated Valve Common Mode Failures During Plant Transients Due to

Improper Switch Settings." The licensee's initial response to the

bulletin was' dated May 14, 1986. The NRC Office of Inspection and

Enforcement is conducting a technical review of this response.

The.NRC. inspector' reviewed the licensee's program for complying with

IEB 85-03. This program is contained in Work Plan 1409.56 and coordinates

IEB 85-03 activities with other motor-operated valve activities such as

environmental qualification (EQ) inspections and preventive maintenance.

The NRC inspector also reviewed the differential pressure testing

procedure (1409.14), the MOVATS testing procedure (1403.31), and the EQ

inspection procedures (1407.37 and 1407.38).

M0 VATS testing was observed for 2CV-5128, 1025, 0711, and 0789. EQ

inspections were observed for 2CV-1511 and 1039. These activities were

found to be conducted in accordance with the approved procedures. The NRC

inspector will observe more testing and inspection activities on

motor-operated valves during the next inspection period.

No violations or deviations were identified.

8. Review of Transient Reports (Unit 1)

ANO-1 had 12 transients in 1985. Six transients were automatic trips, two

were manual trips soon after the events which required plant shutdown, and

eleven transients were the result of malfunctions in the main feedwater

system. All transient reports were reviewed. 'The reports reviewed were

unofficial copies. The formal reports which were on microfilm were not

reviewed. However, the inspector was assured that the reports which were

reviewed were, for the most part, the same as the formal reports. The

following observations were noted:

a. Most reports of transients resulting in plant trips had good plots of

, plant parameters.

I b. One report had a typed page of the sequence of events. Most reports

had a computer printout of the sequence of events. One report did

l not have a sequence of events.

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c. Two events involved manual initiation of HPI. This is considered

significant but was not discussed in the description of the event.

d. To control the transients to preclude an automatic trip, the pilot

operated relief valve (PORV) and pressurizer spray were operated

during some events. This is considered significant but was not

always discussed in the description of the event.

e. It was difficult to identify the plant parameters in the safety

parameter display system (SPDS) computer printouts because the

parameters were not identified by name. Also, the SPDS plotted data

was difficult to read.

No violations or deviations were identified.

9. Review of the 10 CFR 50.59 Determinations of Procedure Changes and Design

Change Packages (Units 1 and 2)

The procedure for Procedure Review, Approval and Revision Control,

1000.06, Revision 22 was reviewed as it applied to the requirements for

the 50.59 determination. The procedure provided for a 50.59 review and

determination by completion of a one page form 1000.06E consisting of

seven questions with boxes of "Yes" or "No" to check. The seven questions

did relate to the three factors of the determination for an unreviewed

safety question as required by 10 CFR 50.59. There was a small space for

a narrative discussion for the basis for the determination. There were no

instructions for completing the form. Also, there was no indication for

the need for a written safety evaluation for the basis of the

determination.

The procedure for Design Control, 1032.01, Revision 7 was reviewed.

Attachment 7 of the procedure provided guidelines for preparation of a

safety and environmental determination which included a 50.59

determination. The guidance appeared to be very complete; however, the

determination is recorded on a Form 202F9 which provided for a check off

for the three factors for determining that an unreviewed safety question

was not involved. The form provided for a brief justification for the

"No" answers given in check off. It also provided for the identification

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of the FSAR sections involved. Although the guidance provided a

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discussion for the need of a written safety evaluation, the Form 202F9

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provided no space specifically for a safety evaluation. It is implied

I. then that the Form 202F9 is the Safety Evaluation.

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The procedure and a change to the procedure for Motor Driven Emergency

Feedwater Pump Test, 1311.04, Revision 0 was reviewed. The change package

l provided a Form 1000.06E. All seven boxes to the questions were checked

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l off as "No." The " Basis of Determination" included a brief description of

how and why the test was to be performed. The determination was not

explicit on why the procedure or change was not an unreviewed safety

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question. The three factors for making the 50.59 determination were not

discussed and there was no written safety evaluation for the

determination.

The following design change packages (DCPs) were reviewed for the 50.59

determinations:

a. DCP 81-1080A, Modification of Diesel Generator Voltage Shutdown

Device -- Electrical

b. DCP 83-1007, Provide Disconnect Switch for "B". Service Water Pump

c. DCP 83-1012, Provide Isolation Device for Pressurizer ERV Block Valve

d. DCP 83-1107, Acoustic Valve Monitoring System Replacement

e. DCP 83-1170, LPI Flow Indication, HPI Flow Indication, Containment

Spray Flow

For each DCP Form 202F9 was completed with a brief justification for the

"No" checks (answers to questions which address the three factors which

must be considered in the determination if the change is an unreviewed

safety question). There were no written safety evaluations which provided

bases for the determinations. Also, in the " Justifications," each of the

three factors necessary for the determination that there were no

unreviewed safety questions were not explicitly addressed. Revisions to

the DCPs did not have 50.59 determinations.

Failure to provide written safety evaluations which provide the bases for

the determination that the above test and design changes did not involve

an unreviewed' safety ' question is an apparent violation. (313/8621-01)

The DCPs did have very detailed descriptions of the changes involved.

These descriptions could be good foundations to 50.59 determinations, but

were not used for this purpose. The Form 202F9 did address the Safety

Analysis Report sections affected.

The licensee recognizes weaknesses in the 50.59 determinations and has

started a program to improve this area. A policy paper is under

preparation which addresses the training and guidance which, for the most

part, is provided in Attachment 7 of the Procedure for Design Control.

10. Refueling Activities (Unit 2)

The purpose of this area of inspection was to ascertain whether refueling

activities are being controlled and conducted as required by Technical

Specifications (TS) and approved procedu es.

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The NRC inspector reviewed the following licensee procedures and compared

them to Section 9 of the TS. It appeared that the procedures implemented

the TS properly.

2502.01 " Refueling Shuffle"

'2502.03- " Preparation for Refueling"

1015.038-8 " Mode 6 Log"

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Thr'oudhroutinedailyobservationsandspecificobservationsoffuel

movement during core off load from both inside the reactor building and

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the spent fuel area of the, auxiliary building, the NRC inspector concluded

', that the licensee appeared to be conducting refueling activities in

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faccordance wit,h the approved procedures.

. No violations or' deviations were identified.

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11., Exit Interview

The NRC inspectors inet with Mr. E. Ewing, Acting Director, Site Nuclear

__ ~ Operations, and other members of the AP&L staff at the end of this

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

inspection and the' findings.