ML20151R173
| 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: |
| 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.
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U. S. NUCLEAR REGULATORY C0FNISSION
REGION IV,
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. Inspection Report: 50-313/88-20
Licenses: DPR-51
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50-368/88-20
Dockets: 50-313
50-368
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Licensee: Arkansas Power & Light. Company
P. O. Box 551
~ ittle Rock, Arkansas 72203
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Facility Name: Arkansas Nuclear One (AN0), Units 1 and 2
' Inspection At: AN0 Site, Russellville, Arkansas
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Inspection Conducted:
June 1 through June 30, 1988
Inspectors:
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7/6/88
W. D. Jot)ay6n, Senior Resident Reactor
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7/ 7/89
R. C. Haag, Resident Reactor Inspector
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Approved:
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D. D. Chamberlain. ~ Chief, Reactor Project
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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
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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
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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
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- J. Vandergrift Operations Manager
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.
-3.
Operational Safety Verification (71707, 71709, 71710, and'71881) ('Jnits 1
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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)
Monthly test of Emergency Diesel Generator 2K4A (Procedure 2104.36,
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Supplement 1).
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Semi-annual test of Unit I containment escape air lock for overall
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
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
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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
the establishment of measures to assure that applicable regulatory
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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)
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The NRC inspector reviewed the licensee's response concerning the analysis
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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