ML20151R173

Insp Repts 50-313/88-20 & 50-368/88-20 on 880601-30. Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint,Surveillance & Temporary Instructions

ML20151R173
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/28/1988
From:	Chamberlain D, Hagg R, Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20151R165	List: ML20151R160 ML20151R167 ML20151R173
References
50-313-88-20, 50-368-88-20, NUDOCS 8808120080
Download: ML20151R173 (10)

See also: IR 05000313/1988020

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= APPENDIX B. 1

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U. S. NUCLEAR REGULATORY C0FNISSION .

REGION IV,

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. Inspection Report: 50-313/88-20 Licenses: DPR-51

50-368/88-20 NPF-6

Dockets: 50-313

50-368

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Licensee: Arkansas Power & Light. Company

P. O. Box 551

~L ittle Rock, Arkansas 72203

Facility Name: Arkansas Nuclear One (AN0), Units 1 and 2

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' Inspection At: AN0 Site, Russellville, Arkansas

Inspection Conducted: June 1 through June 30, 1988

Inspectors: bl//) 7/6/88

W. D. Jot)ay6n, Senior Resident Reactor D' ate'

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InspectoV

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R. C. Haag, Resident Reactor Inspector

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Approved: / >

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D. D. Chamberlain. ~ Chief, Reactor Project Date

Section A, Division of Reactor Projects

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' Inspection Summary

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Inspection Conducted June 1-30, 1988 (Report 50-313/88-20)

_ Areas Ins _pected: Routine, unannounced inspection including-operational safety

verification, maintenance, surveillance, and temporary instructions. :

Results: Within the four areas inspected, two violations were identified

(failure to provide timely corrective action, paragraph 3; and failure to

properly control the ' design criteria of a plant modification, paragraph 5).

Inspection Conducted June 1-30,'1988 (Report 50-368/88-20)

Areas Inspected: Routine, unannounced inspection including operational safety

verification, maintenance, surveillance, temporary instructions.

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

(failure to provide timely corrective action, paragraph 3).

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DETAILS

1. Persons Contacted

AP&L-

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• J. Levine. Executiva Director, AND Site Operations

W. Butzlaf_f, Quality Assurance Supervisor

• A. Cox, Unit 1 Operations Superintendent <

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E. Ewing, General Manager, Technical Support

L. Gulick, Unit 2 Operations Superintendent

C. Halbert, Engineering _ Supervisor

D. Harrison, Plant Engineer

D. Howard, Licensing Manager

• L. Humphre,o, General Manager, Nuclear Quality

G. Kendrict, I&C Maintenance Superintendent

• R. Lane, '.ngineering Manager

• D. Loma , Plant Licensing Supervisor

A. McGrt: gor, Engineering Services Supervisor

• J. McWilliams, Maintenance Manager

• P. Michalk, Licensing Engineer

V. Pettus, Mechanical Maintenance Superintendent

D. Provencher, Quality Assurance Supervisor

• S. Quennoz, General Manager

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P. Rehm, Mechanical Maintenance Engineer

l C. Shively, Plant Engineer Superintendent

C. Taylor, Unit:2 Operations Technical Support Supervisor

J. Taylor-Brown, Quality Control Superintendent

L. Taylor, Special Projects Coordinator

J. Teeter, Operations Technical Support

R. Tucker, Electrical Maintenance Superintendent

l *J. Vandergrift Operations Manager

i C. Zimmerman, Unit 1 Operations Technical Support Supervisor

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• Present at exit interview.

The NRC inspectors also contacted other plant personnel, including

operators, technicians, and administrative personnel.

2. Plant Status (Units 1 and 2)

Unit 1 operated at near 85 percent power and Unit 2 at 100 percent power

throughout the month of June 1988.

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-3. Operational Safety Verification (71707, 71709, 71710, and'71881) ('Jnits 1

ano 2)

The NRC inspectors observed control room operations. reviewed applicable

logs, and condut.ted discussions with control room operators. The NRC

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inspectors verified the operability of selected emergency systems,

reviewed tag-out records and verified proper return to service of affected

components, 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 in accordance

with the station security plan. The NRC inspectors verified

implementation of radiation protection controls during observation of

plant activities.

The NRC inspectors toured accessible areas of 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.

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

electrical system to verify operability. The walkdown was conducted using

various checklists of licensee Procedure 1107.01, "Electrical System

Operations," Revision 28. During the walkdown inspection, the following

items were identified. The licensee was informed of these items so

corrective action could be taken.

Procedure 1107.01 lists the desired position of spare Breaker 5733 as

open, however, the breaker was actually closed. The licensee stated

the breaker will be opened.

The description of Breaker H-25 in Procedure 1107.01 and on the

breaker label was incorrect. The description should be "Startup

Transformer #1 Supply to H-2" in lieu of "Startup Transformer #1

Supply to H-1." The licensee has corrected the breaker labeling

which was an apparent typographical error.

The descriptions for Breakers 24 and 52 on Panel Y01 and Breakers 24

and 52 on Panel Y02 in Procedure 1107.01 did not match the breaker

label description. The licensee has stated the description in the

procedure will be revised for clarity.

125 VDC Panel D11 had several breakers with two different breaker

numbers attached to the individual breakers. The correct breaker

numbers were listed but in some cases an additional outaated number

was also listed. The licensee has corrected the breaker numbering.

The description in Procedure 1107.01 for Breaker 11 on Panel RS-1

listed two different loads for the breaker; however, the breaker

label lists only one load. The licensee has stated the breaker label

will be updated to reflect both loads.

• Breaker No. 41 on Panel Y02 is listed as a spare with the desired

position as open in Procedure 1107.01. The breaker was closed with

the label description, T/C Signal Converter T-08A. The licensee has

stated the breaker label will be corrected to reflect the breaker if

a spare, and that the breaker will be opened.

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In Procedure 1107.01 the description of Breaker No. 5531 included the

word, (Future), and the desired position was listed as open. The

breaker is actually closed. The licensee had a pending revision to

this procedure to list the desired position as closed.

Similar minor examples of procedural / labeling deficiencies are noted in

NRC Inspection Report 50-368/88-15. The licensee has an ongoing

corrective action program in place to correct the kind of

procedural / labeling deficiencies identified above. The NRC inspector

noted that a recent change (Revision 33) to Procedure 1015.01, "Conduct of

Operations," now provides instructions to identify and document

procedural /lateling deficiencies during system alignments. While no

violation will be cited at this time for the identified minor

deficiencies, the effectiveness of the licensee's ongoing corrective

action program will be monitored during future NRC inspections.

The NRC inspector also observed caution cards dated June 19, 1986,

attached to breakers 0123 and 0124 on DC Bus, D01. The caution cards

stated the breakers could not be operated with the outer breaker lever.

To change the position of these breakers, an operator would have to open

the front panel cover and move the actual breaker switch. In addition,

Procedure 1203.02, "Alternate Shutdown," requires breaker 0124 to be

opened under conditions that require plant shutdown outside the control

room. During further review, the NRC inspector discoverd that the

inability to operate the breakers with the outer breaker lever resulted

f rom replacement of the breaker. Design Change Package 83-1032 which

replaced breakers 0123 and 0124 was completed on April 14, 1985.

The licensee stated the delayed time period for correction of the

deficiencies was attributed to incomplete repair parts during the last

repair effort and the required scheduling of repairs during reactor

shutdown. However, the NRC inspector was concerned with the excessive

amount of time elapsed without these breakers being repaired, particularly when

considering breaker 0124 will require operation during alternate reactor

shutdown. Failure to promptly correct the breaker deficiency is an

apparent violation (313/8820-01).

These reviews and observations were conducted to verify that facility

operations were in conformance with the requirements established under

Techniccl Specifications, 10 CFR, and administrative procedures.

4. Monthly Surveillance Observation (61726) (Units 1 and 2)

The NRC inspector observed the Technical Specification required

surveillance testing on the Unit 2 Emergency Diesel Generator 2K48

(Procedure 2104.36, Supplemeat 2) and verified that testing was perfonned

in accordance with adequate procedures, that test instrumentation was

calibrated, that limiting conditions for operation were met, that removal

and restoration of the affected components were accomplished, that test

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results conformed with Technical Specifications and procedure

requirements, and that test results were reviewed by oersonnel other than

the individual directing the test.

During the diesel. generator run, the NRC inspector noted high fuel oil

pressure in the range of 28-30 psi as indicated on gauge 2PI-2987A/8. The

operator's logsheet provided the acceptable range of fuel oil pressure as

10-20 psi. The technical manual for diesel generators specifies a

pressure of approximately 15 psi being built up and maintained in the fuel

supply line. Job Order 728294 was issued in ~ January 1987; however, due to

replacement parts being incorrectly ordered, the repairs have not been

completed. The NRC inspector questioned.if an operability assessment had

been performed for adverse effects associated with the higher than normal

fuel oil pressure. The licensee could not provide evidence that the

assessment had been performed concerning the higher than normal fuel oil

pressure and effects on diesel generator operation. ~ After subsequent

discussions with the vendor, the licensee stated tnat fuel oil pressure in

the range of 28-30 psi was acceptable for extended operations. The

decision was made on the basis of an initial hydrostatic test of the fuel

oil system to 50 psi and the vendor recommendation that fuel oil pressure

of 30 psi was acceptable for diesel generator operation. Failure to

promptly repair the high fuel oil pressure and to properly evaluate the

significance of the high pressure is a second example of the violation in

paragraph 3 (368/8820-01).

The NRC inspector also witnessed portions of the follos.Ing test

activities:

Semi-annual test of Unit 2 containment personnel air lock for overall

air leakage (Procedure 2304.022, Job Order 758018). The initial test

provided an unacceptable leak rate of 9700 cc/ minute.- Following

repairs (seemaintenancesection)asubsequentleaktestwas

satisfactorily performed.

• Monthly test of diesel fuel from Emergency Diesel Fuel Tank T578

(Procedure 1618.010). The diesel fuel was checked for viscosity,

water and sediment.

• Monthly test of Emergency Feedwater Pump 2P78 (Procedure 2106.006,

Supplement II). During the surveillance an equalization valve for

Test Gage 2FI-0798A was repositioned. The valve was not labeled.

The NRC inspector noted that possible confusion could exist in

identifying this equalization \sive and an additional equalization

valve for an adjacent gauge that was labeled "Equalization Valve."

The licensee has subsequently labeled the equalization valve for

Gauge 2FI-0798A. The NRC inspector also noted the procedure did not

provide instructions for opening or closing the isolation valves for

the local suction and discharge pressure gauges. The licensee is

revising the procedure to provide operating instructions for the

isolation valves.

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Monthly surveillance of Channel B Excore Instrumentation

(Procedure 2304.101, Job Order 758015)

Monthly test of Channel D of emergency feedwater initiation and

controlsystem(Procedure 1304.148, Job Order 759229)

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Monthly test of Emergency Diesel Generator 2K4A (Procedure 2104.36,

Supplement 1).

Semi-annual test of Unit I containment escape air lock for overall

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air leakage (Procedure 1304.020, Job Order 768010). The NRC

inspector noted an electrical terminal box that was not mounted to

the outer face of-the air lock barrel. The terminal box was

associated with the interlock features of the air lock. Job

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Order 760439 was issued to remount the terminal box and Condition

Report 1-88-055 was written to determine the cause of the box not

being mounted.

In addition, regional NRC inspectors witnessed portions of the following

activities:

Quarterly pressurizer level response test (Procedure 2103.05)

Monthly control room emergency air condition system test

(Procedure 2104.07)

No additional violations or deviations were identified.

5. Monthly Maintenance Observation (62703) (Units 1 and 2)

Station maintenance activities for the 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 or standards; and in conformance with the Technical

Specifications.

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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 were inspected

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

to returning components or systemr. 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 the status of outstanding jobs

and to ensure that priority is assigned to safety-related equipment

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maintenance which nay affect system performance.

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The following maintenance activities were observed / reviewed:

Troubleshooting the dual position indication from CV-3840 limit

switches (Job Order 758527). The lower limit switch was determined

to be defective. The licensee initiated procurement of the

replacement part.

Fabrication and installation of a stem position indication rod on

CV-1407 (borated water storage tank outlet valve) (Job Order 758205).

Tube leak repairs on Drain Heater E8A (Job Order 758501). While this

component is not safety-related, the NRC inspector observed portions

of the repair and the isolation of the drain heater which required

bypassing a portion of one feedwater train and reduction of reactor

power to 60 percent.

Investigation of air leckage on the Unit 2 personnel air lock (Job

Order 3005.) Following the tightening of several plugs and packing

adjustment on the hatch handwheel the subsequent air leakage test was

performed satisfactory.

Replacing upper motor bearing on Service Water Punp 2P-4A

(Procedure 2403.04, Job Order 759142)

Troubleshooting core protection calcul6 tors Channel A (Job

Order 758677)

Temporary modifications performed on the reed switch position

transmitter signal for control element assemblies (CEA) Nos. 28

and 66 (Job Orders 759874 and 759945). This modification provides a

continuous rod full out signal from one of the two reed switch

position transmitters for each CEA. Prior to this modification the

position transmitters were repeatedly providing spurious and

incorrect output signals.

Packing repair of CV-2617 (isolation valve in supply steam line to

emergency feedwater pump turbine) (Job Order 759073). Sealant was

injected into the packing area to stop a steam leak.

Procedure 1025.015 "On Line Repair Procedures" which was referenced

in the job order, does not have a specific task which correlates with

the method used for sealant injection. This and several other minor

comments concerning the procedure were identified to the licensee.

Troubleshooting the failed high Steam Generator B high range level

indication (JobOrder 759835). Transmitter LT-2673 which provides

input to Channel B of the emergency feedwater initiation and control

system was determined as the cause of the failed high level

indication. Due to the location of LT-2673 in the reactor building,

the licensee evaluated various accident senarios with LT-2673 failed

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high and concluded continued operations with LT-2673 failed high was

justified. The licensee has prepared plans to investigate LT-2673

failure during the next unit shutdown.

Service Water Check Valve SW-1A heat leakage repair (Job

Order 797184). Due to the extent of repairs, Plant Change 88-1919

was issued for the plant modification involving parts replacement. A

new hinge pin with a larger diameter was installed to reduce the

excessive movement of the joint. The new pin was manufactured from

304 stainless steel barstock with a yield strength of 30 ksi. The

original hinge pin was 416 stainless steel with a yield strength of

40 ksi. The engineering justification used for the decrease of yield

strength was based on increasing the new pin diameter 1/8 inch. This

would result in a cross sectional area increase of 26 percent that

would offset the reduced yield strength.

The fabrication instruction in the plant change did not specify a

1/8-inch increase in pin diameter but required the new pin be

machined to fit the smallest dimension of the mating parts. During

observation of the repair, the NRC inspector questioned the machinist

on the actual increase in pin diameter and learned the pin was

increased only 1/16 inch. Later the NRC inspector questioned the

plant engineer if he had received information on the actual increase

in pin diameter and taken steps to modify the engineering

justification used for the decrease in yield strength. The engineer

had not received notice from the field concerning a change in the

basis (diameter increase of 1/8 inch) used in the engineering -

justification nor did the plant change have a means for identifying

this information. Following the questioning by the f;RC inspector, the

licensee performed a detailed calculation to verify the new pin was

acceptable. The NRC inspector was concerned with the broader

implication of this modification in that a change to the basis used

in the justification of a design change, if not specifically

delineated in the plant change, may not be identified to the

engineer. The NRC inspector reviewed Procedure 1032.01, Plant

Engineering Action Requests and Plant Changes, and Procedure 1032.02,

Installation Technical Support. These procedures provide

instructions for the plant change process, including the closeout

process; however, they do not provide instructions to ensure a change

to a basis in the justification of a design change is properly

identified. This procedural concern is an apparent violation of

10 CFR 50, Appendix B, Criteria III, Design Control, which requires

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the establishment of measures to assure that applicable regulatory

! requirements and design basis are correctly translated into

specifications, drawings, procedures and instructions (313/8820-02).

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6. Verification of Changes Made to Comply With PWR Moderator Dilution

Requirements (Temporary Instruction 2515/94) (Unit 1)

f The NRC inspector reviewed the licensee's response concerning the analysis

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of the potential for, and the consequences of, a boron dilution accident

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for Unit 1. Only one situation was identified in which a single valve

failure would allow a boron dilution accident due to Na0H injection in the

reactor coolant system. This event could have occurred when filling the

refueling canal by way of the low pressure injection system and when

testing the Na0H control valves. In this response the licensee stated

that operating procedures had been changed to require an additional valve

to be manually closed during Na0H control valve testing. The NRC

inspector verified that Supplement II to Procedure 1104.005, Reactor

Building Spray System Operation, requires valve CA-49 be closed when

testing the Na0H control valves if using the low pressure injection system

to fili the refueling canal.

No violations or deviations were identified.

7. Verification of Quality Assurance (QA) Regarding Diesel Generator (DG)

Fuel Oil (Temporary Instruction 2515/93) (Units 1 and 2)

The NRC inspector reviewed the licensee's QA Manual for operations to

determine if DG fuel oil is included in the QA program. While DG fuel oil

is not listed on the summary Q-lists that are located in the safety

analysis reports or on the component level Q-lists, quality assurance is

pruvided under the controls of expendable and/or consumable items. These

controls verify compliance with Technical Specifications and additional

standards identified by the licensee. The NRC inspector reviewed the

following procedures:

1618.010 Sampling Diesel Fuel ANO-1

1618.035 Diesel Fuel Oil Transport Sample

2618.005 Sampling Diesel Fuel AN0-2

In addition to these required tests, the licensee has initiated quarterly

testing of DG fuel oil for compliance with the DG vendor recommended fuel

oil requirement. The licensee has also implemented a program of recycling

and filtering the fuel oil in the emergency diesel fuel oil tanks on an

18-month basis. QA involvement with these activities is ccmmensurate with

other safety-related activities. As a result of this inspection, the NRC

inspector found that the licensee has included DG fuel oil in the QA

program.

No violations or deviations were identified.

8. Exit Interview

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

Operations, and other members of the AP&L staff at the end of th~e

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

inspection and the findings.

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