ML20206S079
| ML20206S079 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/01/1986 |
| From: | Bruce Bartlett, Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20206S054 | List: |
| References | |
| 50-482-86-08, 50-482-86-8, NUDOCS 8607070286 | |
| Download: ML20206S079 (9) | |
See also: IR 05000482/1986008
Text
-
. _ -
..
.
APPENDIX B
US NUCLEAR REGULATORY COMMISSION
Region IV
4
NRC Inspection Report:
50-482/86-08
License: NPF-42
Docket: 50-482
Licensee: . Kansas Gas and Electric Company (KG&E)
Post Office Box 208
Wichita, Kansas 67201
Facility Name: Wolf Creek Generating Station (WCGS)
Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas
a
Inspection Conducted: April 1-30,1986
j
h
7I
(>
'
Inspector-
&
B. L. Bartlett, Resident Reactor Inspector,
Date
Operations
.
6)
!
'
r
I
/
l
Approved: 3 .
,
\\ R. @ nter,' Chief, Project Section B,
Date
J
Reactor Projects Branch
Inspection Summary
Inspection Conducted April 1-30, 1986 (Report 50-482/86-08)
Areas Inspected: Routine, unannounced inspection including followup on a
previously identified NRC item; operational safety verification; engineered
safety features system walkdown; onsite followup of events; monthly
surveillance observation; monthly maintenance observation; and plant tours.
Results: Within the seven areas inspected, three violations were identified
(failure to verify correct revision of a surveillance procedure prior to use,
paragraph 6; failure to maintain auxiliary building fire barrier penetrations
in accordance with TS, paragraph 8; and failure to adequately perform TS
surveillance requirement, paragraph 7).
kDj70gCK05000482 ' 'h!
e6 860702
G
, _ f)
,
-
>--
r
,
--
-
--,,w
ow
.
_2
DETAILS
1.
Persons Contacted
Principal Licensee Personnel
- J.
A. Bailey, Interim Site Director
- G. P. Rathbun, Manager, Engineering and Technical Services Support
- F. T. Rhodes, Plant Manager
- R.
M. Grant, Director, Quality
M. Estes, Superintendent of Operations
M. D. Rich, Superintendent of Maintenance
- M. G. Williams, Superintendent of Regulatory, Quality, and
Administrative Services
0. L. Maynard, Manager, Licensing
K. Peterson, Licensing
- H. Chernoff, Licensing
'
- G. Pendergrass, Licensing
..
- W. M. Lindsay, Supervisor, Quality Systems
- C. J. Hoch, QA Technologist
- A. A. Freitag, Manager, Nuclear Plant Engineering-WCGS
- C. M. Herbst, Project Engineer, Bechtel
M. Megehee, Compliance Engineer
'
- W. J. Rudolph, QA Manager
- R. L. Hoyt, Acting Operations Coordinator
The NRC inspector also contacted other members of the licensee's staff
during the inspection period to discuss identified issues.
- Denotes those personnel in attendance at the exit meeting held on
May 7, 1986.
2.
Followup on Previously Identified NRC Items
(Closed) Open Item (482/8511-17):
Submit Information Regarding NRC Generic Letter 83-28
This item concerned a licensee commitment to submit specific information
regarding the periodic maintenance program and trending of parameters for
the reactor trip breakers. The information committed to was submitted by
the licensee on July 17, 1985. This item is closed.
3.
Operational Safety Verification
The NRC inspector verified that the facility is being operated safely and
in conformance with regulatory requirements by direct observation of
licensee facilities, tours of the facility, ir)terviews and discussions
with licensee personnel, independent verification of safety system status
and limiting conditions for operations, and by reviewing facility records.
.
-3-
The NRC inspector, by observation, interview of personnel, and review of
.
documents, verified that physical security was being implemented in
accordance with the' site security plan and that radiation protection
activities were controlled.
By observing selected valve position, electrical breaker position, and
control room indication, the NRC inspector confirmed the operability of
the residual heat removal system and the auxiliary feedwater system.
No violations or deviations were identified.
4.
Engineered Safety Features (ESF) System Walkdown
The NRC inspector verified the operability of selected ESF systems by
walking down selected accessible portions of the systems. The NRC
inspector verified selected valves and electrical circuit breakers were in
the required position, power was available, and valves were locked where
required.
The NRC inspector also inspected system components for damage
or other conditions that could degrade system performance.
The ESF system listed below was walked down during this inspection
report period:
o
Essential Service Water (EF)
The NRC inspector utilized the following documents during the EF
walkdown.
M-02-EF-01,
Emergency. Service Water
P&ID (Q), Rev. 10
M-02-EF-02,
Emergency Service Water
P&ID (Q), Rev. 12
M-KR-EF-01,
Emergency Service Water
P&ID (Q), Rev. 14
Procedures Checklist,
CKL-EF-120, EF Valve
Breaker and Switch Lineup, Rev. 8
No violations or deviations were identified.
5.
Onsite Followup of Events
The NRC inspector performed onsite followup of nonemergency events that
occurred during this report period.
The NRC inspector (when available)
-observed control room personnel response, observed instrumentation
indicators of reactor plant parameters, reviewed logs and computer
printouts, and discussed the event with cognizant personnel.
The NRC
inspector verified the licensee had responded to the event in accordance
1
..m
-
. _ _ _
<
,
'
,
.
,
...t i
-
,
,
,
- .
,.
e
-'
~
"'~
"
'
-4-
"
e
with procedures and had notified the NRC and other agencies as required in
a timely fashion.
Engineered safety feature actuations that occurred during the report
period are listed in the table below.
Where applicable, the NRC inspector
will review the LER for each of these events and will report any findings
in subsequent NRC inspection reports.
Date
Event *
Plant Status
Cause
4/8/86
Mode 1
Personnel Error
Rx Trip
4/10/86
Rx Trip /AFAS/
Mode 5
S/G Level
MFWIS
Spiked Low
4/10/86
CPIS
Mode 5
Inadequate Test
Points'in
Procedure
,
4/10/86
CPIS/CRVIS
Mode 5
Radiation Monitor
Sample Pump
Blown Fuse
4/13/86
CRVIS
Mode 5
Chlorine Monitor
Lost Power
4/20/86
FWIS/ Turbine
Mode 3
S/G Swell
Trip
- Event
CRVIS - control room ventilation isolation signal
CPIS - containment purge isolation signal
FWIS - feedwater isolation signal
AFAS - auxiliary feed water actuation signal
i
Rx Trip - Reactor Trip
,
MFWIS - main feedwater isolation signal
.
{
Selected NRC inspector observations are-discussed below:
o
The reactor trip which occurred on April 8, 1986, was caused by KG&E
meter and relay personnel working in the 345 KV switchyard.
Two
terminals were inadvertently shorted while connecting a test
-
instrument, resulting in'a transformer primary differential
protection actuation.
A main generator trip, a turbine trip, and a
j
reactor trip (power was greater than 50%) occurred. The NRC
l
inspector was onsite during the trip and responded to the control
room when he heard the plant announcement.
Control room personnel
,
were observed to respond to the event in accordance with procedures
!
l
and maintained positive control of the plant at all times.
During
the recovery period immediately following the trip, the supervising
I
!
l
!
l
'
_
_
.
,
'.
i
-5-
l
l
operator (50) was observed to be making use of the safety parameter
display system (SPDS).
The safety assessment system (SAS) indicated
some critical safety functions (CSF) trees as " red path", (a path
needing further evaluation) due to the operator throttling the
auxiliary feedwater flow to control primary system average
temperature while recovering normal steam generator levels.
The
inspector observed that, on the whole, the SAS was useful.
All
equipment operated as designed.
No violations or deviations were identified.
,
6.
Monthly Surveillance Observation
The NRC inspector observed selected portions of the performance of
surveillance testing and/or reviewed completed surveillance test
procedures to verify that surveillance activities were performed in
accordance with TS requirements and administrative procedures.
The NRC
inspector considered the following items while inspecting surveillance
'
activities:
J
o
Testing was being accomplished by qualified personnel in accordance
with an approved procedure.
o
The surveillance procedure conformed to TS requirements.
o
Required test instrumentation was calibrated.
o
TS limiting conditions for operation (LCO) were satisfied.
o
Test data was accurate and complete. Where appropriate, the NRC
inspectors performed independent calculations of selected test data
3
!
to verify their accuracy.
o
The performance of the surveillance procedure conformed to applicable
administrative procedures.
o
The surveillance was performed within the required frequency and the
test results met the' required limits.
Surveillance Tests (STS) witnessed and/or reviewed by the NRC inspectors
are listed below-
,
o
STS RE-004, Revision 3, " Shutdown Margin Determination;" Xenon
Free, 200 F; Control bank only in.
,
2
o
STS RE-004, Revision 3, " Shutdown Margin Cetermination;"
Current conditions, shutdown banks out and reactor coolant
system boron concentration (C ) 523 ppm.
g
STSRE-004,Revfsion3,"ShutdownMarginDetermination;"
o
No Xenon, 200 F and all rods in (ARI).
,
4
+..
- - -
r
y
. e
-
-
.
,,,m
--,.
y
.
.
-6-
o
STS KJ-001A, Revision 2, " Integrated D/G and Safeguards Actuation
Test-Train A"
o
STS KJ-001B, Revision 2, " Integrated D/G and Safeguards Actuation
Test-Train B"
o
STS BB-011, Revision 4, "RCS and Pressurizer HU/CD Surveillance"
o
STS NB-005, Revision 3, " Breaker Alignment Verification"
STS EP-210, Revision 5 "ECCS Accumulator Inservice Check Valve
o
Test"
Selected NRC inspector observations are discussed below:
On April 8,1986, during the review of STS RE-004, " Shutdown Margin
o
Determination," the NRC inspector observed that even though the
surveillance had been signed off by the shift supervisor as complete,
that the test performer had, contrary to procedure, neglected to
verify and to document that the procedure being used was the correct
revision and/or that temporary changes were current. This is an
,
I
apparent violation (482/8608-01).
o
On April 8,1986, the NRC inspector manually performed a shutdown
margin determination using the licensee's procedure, and compared the
'
results with the answers generated by the licensee's computer
program. The NRC inspector's results agreed with the licensee's
computer results within expected error margins.
o
During this report period, STS KJ-001A and STS KJ-001B were inspected
to ascertain whether functional testing of a complex safety-related
system and related subsystems were in conformance with regulatory
requirements and industry guides or standards referenced in the
application and other correspondence to the NRC. Selected portions
of the surveillances were observed by the NRC inspector in addition
to reviewing selected portions of the documentation. Prior to the
surveillance performance, the NRC inspector reviewed the test
procedures and, except for some minor typographical errors and
editorial suggestions, had no coments. During the performance of
STS KJ-001A, three events occurred which the NRC inspector followed
up:
Approximately 1 minute af ter Step 5.2.17 was performed, the ' A'
a.
Train ESF Status Panel SA-066-X went dark when its fuse blew.
This did not affect the performance of the test. When the fuse
was replaced, the blown fuse was verified to be the correct
size. During the performance of STS KJ-001B, its status panel
did not blow a fuse. The fuse failure of SA-066-X was
considered to be random.
.
-.
._
.
.
-7-
b.
ESH Pump House Heater EGD01C, Breaker NG05EHF3, did not
"loadshed" on Step 5.2.21 due to a ground.
c.
Component Cooling Water Pump No. 3, PEG 01C, did not restart
after "loadshed" due to incorrect test-jumper installation.
During the performance of STS KJ-001B, there was one item on which
the NRC inspector performed followup:
d.
Standby AC Lighting XQB51, Breaker NG06EHF2, did not "loadshed"
on Step 5.2.23 as required due to a ground.
The NRC inspector verified that appropriate work request'(WR)'and: plant
modification requests (PMR) were written and retesting performed as
required for all test deficiencies noted above.
7.
Monthly Maintenance Observation
The NRC inspector observed maintenance activities performed on
safety-related systems and components to verify that'these act vities were-
conducted in accordance with approved procedures,;TS, and applicable
industry codes and standards.
The following elements were considered by
the NRC inspector during the observation and/or review of the. maintenance
activities:
,
'
-
o
LC0 were met and, where applicable, re'dundant components were .
-
'
o
Activities complied with adequate administrative controls,
o
Where required, adequate, approved, and up-to-date procedures were
used.
o
Craftsmen were qualified to accomplish the designated task and
technical expertise (i.e., engineering, health physics, operations)
was made available when appropriate.
o
Replacement parts and materials being used were properly certified.
o
Required radiological controls were implemented.
o
Fire prevention controls were-implemented where appropriate,
o
Required alignments and surveillances to verify post maintenance
operability were performed.
o
Quality control hold points and/or checklists were used when
appropriate and quality control personnel observed designated work
activities.
l
t
. ____
_ _ _ -
.
,
.
-g_
Selected portions of the maintenance activities accomplished on the work
requests (WR) listed below were observed and related documentation
reviewed by the NRC inspector:
o
WR 0934-86, " Hydro Test Of AE-V330"
WR 07082-85, " Rework Limitorque Operator and Perform Scheduled
o
Maintenance" (AL HV-033)
WR 07085-85, " Rework Limitorque Operator' and Perform Scheduled
o
Maintenance" (AL hV-032)
NRC inspector findings are discussed below:
,
.
The NRC inspector observed selected portions of WR 0934-86 and verified
the hydro test arrangement was in accordance with procedure MGM M00C-02,
Revision 0.
During a routine plant tour on April 29, 1986, the NRC inspector observed
WR Tag No. 25559 on Valve AL HV-032 and WR Tag No. 25501 on Valve HV-033
(essential service water to the turbine driven auxiliary feedwater pump).
These WRs were written to document that these two valves failed to meet
their specified limiting value of full-stroke time (15 seconds). A WR was
issued and the stroke distance of the valves were shortened, (AL HV-032 to
82 percent and AL HV-033 to 91 percent) then the stroke time surveillance
was reperformed with the valves meeting the 15 second requirement.
WR's 07085-85 and 07082-85 shortened the stroke distance of AL HV-032 and
AL HV-033 on March 5, 1986. However, it was not until April 2,1986, that
Engineering Evaluation Request (EER) 86-AL-01 was written to request
approval from site nuclear plant engineering (NPE) that this change was
appropriate, and it was not until April 29, 1986, that this request was
granted.
During discussions with licensee personnel concerning procedures and
drawings giving guidance on the proper position setting of open limit
switches, it became apparent that confusion existed.
Schematic
Diagram E-03ALO4B(Q), Revision 2, for Valve AL HV-033 provided a limit
switch contact development which could be incorrectly interpreted to mean
that the switch could ' e set anywhere from 80 percent to 100 percent open
o
or could be correctly interpreted to mean that the switch be set at
100 percent open minus a few percent for valve coast (dependent on valve
size and speed). The NRC inspector was concerned that an open limit
switch on a safety-related valve could be adjusted using a diagram similar
to the one on E-03ALO4B(Q) and the flow rate lowered unknowingly below
that which was assumed in the safety analysis report. After continued
discussion about this drawing, the licensee informed the NRC inspector
that the valve stroke should not have been changed using E-03ALO4B(Q),
Revision 2.
The same discussion was also applicable to Valve AL HV-032.
Reducing the stroke-travel from the fully open position for Valves
AL HV-032 and AL HV-033 is an apparent violation (482/8608-03) of the
testing required by TS 4.0.5 which invokes the ASME Section XI provision
requiring full-stroke time measurements.
-
._.
_
_
_
_
_
._ _
.
- .
.
,
'
1
!
'
9-
!
8.
Plant Tours
3
At various times during the course of the inspection period, the NRC
-
inspector conducted general tours of the reactor building, auxiliary
j
building, radwaste building, fuel handling building, control building,
turbine building, and the secured area surrounding the buildings. During
i
the tours, the NRC inspector observed housekeeping practices, fire
l
protection barriers and equipment, and maintenance on equipment. The NRC
,
inspector also discussed various subjects with licensee personnel.
!
Selected NRC inspector observations are discussed below:
During a plant tour on March 26, 1986, the NRC-inspector observed that the
openings on either side of 'C' containment tendon buttress where it
penetrated the 2047' and 2026' elevations-of the auxiliary building were
covered by a steel plate. The NRC inspector questioned the licensee on
the adequacy of these steel plates as fire barriers. On April 8,1986,
i
the licensee informed the NRC inspector that the installation of the steel
plates was in accordance with design, but that the design was inadequate.
The design should have specified the penetratians to be covered with a
j
3-hour fire barrier and they were not. The licensee established hourly
fire watch patrols in accordance with the .TS rcquirements. The licensee
'
informed the NRC inspector that the inquiry into the cause of the design
error revealed that KG&E was originally notified of the design error on
,
!
June 6, 1985, in a letter from the Architect Engineer. Plant modification
j
Request (PRM) 1157 was originated to correct the design error, but the
personnel processing the PMR failed to recognize the TS-issue involved.
!
The NRC inspector verified through field inspections that additional fire
l
walls and substantial distances separated the. impacted safety-related
j
areas. The failure to establish and maintain fire barriers for the
containment tendon buttress openings is an apparent violation
'
(482/8608-02) of TS 3.7.11 which requires-that all fire barrier
penetrations separating safety-related fire areas shall be operable.
9.
Exit Meeting
The NRC inspector met with licensee personnel to discuss the scope and
findings of this inspection on May 7,~1986. The NRC inspector also
,
attended entrance / exit meetings of other NRC region-based inspectors
identified below:
Inspection
Area
Inspection
i
Period
Inspector
Inspected
Report No.
4-1/4-86
R. Caldwell
Security
86-07
j
4-28/
R. Baer
Radiological
86-11
'
5-2-86
Protection
.
-