ML20235M364
| ML20235M364 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/21/1987 |
| From: | Bruce Bartlett, Cummins J, Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20235M308 | List: |
| References | |
| 50-482-87-20, NUDOCS 8710060163 | |
| Download: ML20235M364 (12) | |
See also: IR 05000482/1987020
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
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' REGION IV
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NRC Inspection Report:
50-482/87-20
License:
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Docket:
50-482-
Licensee:
Wolf Creek Nuclear Operating Corporation'(WCNOC)..
P.-0. Box 411
3
.Burlington, Kansas 66839
Facility Name: -Wolf Creek Generating Station (WCGS)
Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas
Inspection Conducted:
August 1-31, 1987
Inspectors:
bduh,
9 C l 8'7
w
J..E.
C~ummins, Senio'
Resident Inspector,
Date
Operations
00h An
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B. L. Bartlett, Redident Reactor Inspector,
Date7
Operations
Approved:
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4!Ef!/7
D. R. Hunter, Chief, Reactor Project
Date
Section B, Reactor Projects Branch
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Inspection Summary
Inspection Conducted August 1-31, 1987 (Report 50-482/87-20)
Areas Inspected:
Routine, unannounced inspection including plant status,
operational safety verification, monthly surveillance observation, monthly
maintenance observation, 10 CFR Part 21 report followup, physical security
verification, radiological protection, engineered safety features system
walkdown, allegations followup, and followup on regional requests for
information.
Results:
Within the 10 areas inspected, one violation was identified (failure
to enter Technical Specification (TS) 3.0.3, paragraph 3).
One open item is
identified in paragraph 11 and one unresolved item is identified in
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paragraph 5.
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DETAILS
1.
Persons Contacted
- B. D. Withers, President
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- R. M. Grant, Vice President, Quality
- J. A. Bailey, Vice President, Engineering and Technical Services
- F. T. Rhodes, Vice President, Nuclear Operations
- G. D. Boyer, Plant Manager
- 0. L. Maynard, Manager of Licensing
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- C. M. Estes, Superintendent of Operations
- M. D. Rich, Superintendent of Maintenance
- M. G. Williams, Superintendent of Regulatory, Quality, and
Administrative Services
- W. J. Rudolph, QA Manager-WCGS
- A. A. Freitag, Manager, Nuclear Plant Engineering (NPE), WCGS
- M. Nichols, Plant Support Superintendent
- K. Peterson, Supervisor of Licensing
G. Pendergrass, Licensing
- W. M. Lindsay, Supervisor, Quality Systems
- C. J. Hoch, QA Technologist
- J. A. Zell, Training Manager
- J. Houghton, Operations Coordinator-0perations
The NRC inspectors 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
September 4, 1987.
2.
Plant Status
The plant operated in Mode 1 during the inspection period.
3.
Operational Safety Verification
The NRC inspectors verified that the facility is being operated safely and
in conformance with regulatory requirements by direct observation of
licensee facilities, tours of the facility, interviews and discussions
with licensee personnel, independent verification of safety system status
and limiting conditions for operations, and reviewing facility records.
The NRC inspectors, by observation of randomly selected activities and
interview of personnel verified that physical security, radiation
protection, and fire protection activities were controlled.
By observing accessible components for correct valve position and
electrical breaker position, and by observing control room indication, the
NRC inspectors confirmed the operability of selected portions of
safety-related systems.
The NRC inspectors also visually inspected safety
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components for leakage, physical damage, and other impairments that could
prevent them from performing their designed functions.
Selected NRC inspector observations are discussed below:
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On August 18, 1987, during a routine plant tour, the NRC inspector
observed licensee security personnel responding to a possible loss of
security in the fuel building. During maintenance to replace a section of
-fire protection piping, a hole through the side of the fuel building had
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been left after a section of piping had been removed. The personnel
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performing the work failed to realize that the opening left with the pipe
removed could be a security problem and a fuel building pressurization
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problem. The NRC inspector reviewed the circumstances and it was
concluded that the hole created neither a security problem nor a TS
problem; however, it did point out the need to consider the functions of
walls and other boundaries during the preparation of work packages,
particularly when the boundaries are being breached during the work
activity.
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A3 proximately one hour after the above observation, the NRC inspector
o) served that Penetration OP152W1317 (auxiliary building to control room)
had been recently worked on; however, with the help of the shift
supervisor (SS) it was determined that it did not violate the control room
pressure boundary. Discussions with the SS the next day revealed that
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work packages were again found to inadequately address the functions of
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walls when they were breached. The SS had all penetration work stopped
until he felt that the wall penetrations were properly addressed.
Later
that day penetration work was allowed to continue.
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Licensee personnel informed the NRC inspector that even if
Penetration OP152S1317 had been completely breached that it would not
matter since that part of the auxiliary building wall was not part of the
control room pressure boundary. The pressure boundary was actually the
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wall of the electrical cable chase which was inside the control room.
On August 20, 1987, during a routine plant tour of the control room, the
NRC inspector observed Door 36171 propped open. Door 36171 is the access
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door to the southeast electrical chase and as noted above is a part of the
control room pressure boundary. The NRC inspector notified the SS who
immediately implemented appropriate administrative controls.
Interviews
with the personnel who had propped the door open to perform penetration
work revealed that the door had been open off and on all morning and
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afternoon but for less than 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> at any one time. The review by the
NRC inspector determined that the licensee inspection activities in the
area of the door were limited to the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the control room pressure boundary violated, the ability of either
train of the control room emergency ventilation system to maintain the
control room at a positive pressure of greater than or equal to .25 inches
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of water was violated. With both trains of the control room emergency
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ventilation system inoperable, the licensee should have entered TS 3.0.3.
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This failure of the licensee to recognize entry into TS 3.0.3 is an
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apparent violation (482/8720-01).
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4.
Monthly Surveillance Observation
The NRC inspectors 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
inspectors considered the following elements while inspecting surveillance
activities:
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Testing was being accomplished by qualified personnel in accordance
with an approved procedure.
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The surveillance procedure conformed to TS requirements.
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Required test instrumenhtion was calibrated,
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Technical Specification 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
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.
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Surveillance witnessed and/or reviewed by the NRC inspectors are listed
below:
o
STS PE-013, Revision 5, " Personnel Air Lock Seal Test," performed on
August 26, 1987
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STS RE-012, Revision 1, "QPTR Determination," performed on August 26,
1987
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STS SE-002, Revision 0, " Manual Calculation of Reactor Thermal
Power," performed on August 26, 1987
o
STS EJ-1008, Revision 1, "RHR System Inservice Pump 'B' Test,"
performed on August 26, 1987
o
STS EG-100A, Revision 2, " Component Cooling Water Pumps A/C Inservice
Pump Test," performed on July 10, 1987
o
STS EG-100B, Revision 2, " Component Cooling Water Pumps B/D Inservice
Pump Test," performed on July 10, 1987
No violations or deviations were identified.
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5.
Monthly Maintenance Observation
The NRC inspector observed maintenance activities performed on
safety-related systems and components to verify that these activities were
conducted in accordance with approved procedures, Technical
Specifications, 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
LCOs were met and, where applicable, redundant components were
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Activities complied with adequate administrative controls,
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Where required, adequate, approved, and up-to-date procedures were
used.
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Craftsmen were qualified to accomplish the designated task and
technical expertise (i.e., engineering, health physics, 0perations)
was made available when appropriate,
o
Replacement parts and materials being used were properly certified,
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Required radiological controls were implemented,
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Fire prevention controls were implemented where appropriate.
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Required alignments and surveillance to verify post maintenance
operability were performed.
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Quality control hold points and/or checklists were used when
appropriate and quality control personnel observed designated work
activities.
Selected portions of the maintenance activities accomplished on the work
requests (WR) listed below were observed and related documentation
reviewed by the NRC inspector:
No.
Activity
WR 02972-87
Piping Line EF-134-HBC-16, UT thickness examination
WR 02892-87
Foam penetration closures, auxiliary building
WR 02888-87
Foam penetration closures, auxiliary
building-inspect and rework as required
Selected NRC inspector observations are discussed below:
On August 21, 1987, the licensee determined by ultrasonic testing (UT)
that Section EF-134-HBC-16 of essential service water (ESW) Train "B"
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piping had less than the allowable minimum wall thickness of .328 inches.
.The lowest thicknesses measured were .125 inches.
This section of piping
was located directly downstream of throttled butterfly Valve' EF V090.
The
licensee has determined that piping downstream of throttled butterfly
valves was subject to a possible high erosion rate.
The licensee
performed the UT on the pipe in accordance with a recently implemented
preventive maintenance erosion / corrosion monitoring program.
On August 21, 1987, when this piping wall was identified to be less than
minimum, in accordance with licensee Procedure ADM 08-212, Revision 0,
" Erosion / Corrosion Monitoring Program," Work Request (WR) 02972-87 was
written to get an engineering evaluation of the minimum wall condition and
to provide instructions for any necessary repairs.
On August 24, 1987, the engineering evaluation determined that the pipe
would not meet seismic and ASME Code compliance requirements.
Based on
this evaluation, the licensee declared the ESW Train "B"
inoperable and
entered TS 3.7.4, which allowed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to either make the component
operable or to shut down the plant.
The section of pipe was repaired
within the 72-hour action of TS 3.7.4 by overlay welding on.its outer
surface which increased the wall thickness to greater than the required
minimum.
During the refueling outage from October-December 1986, this
same section of piping had been repaired to correct a through wall leak.
Activities related to the repairs to ESW piping that was less than the
required minimum wall thickness due to erosion were discussed in NRC
Inspection Report 50-482/87-15, paragraph 5.
Unresolved Item 482/8715-04 was written pending NRC review of whether or
not the licensee should declare a component inoperable immediately upon
determination that it fails to meet ASME Code requirements.
In this
instance ESW Pipe Section EF-134-HBC-16 was determined to be below minimum
wall thickness by the licensee at 3:41 p.m. (CDT) on August 21, 1987, as
documented in the shift supervisor's log.
However, the NRC inspector
determined from discussions with licensee personnel that the engineering
evaluation was not performed until Monday, August 24, 1987.
The
engineering evaluation determined that the section of pipe would not have
withstood upset conditions (seismic event) as required and ESW Train "B"
was declared inoperable.
Pending completion of the NRC review of the question of the operability of
a component that fails to meet ASME Code requirements, this will remain an
unresolved item (482/8720-02) and will be reviewed as a part of Unresolved
Item 482/8715-04.
As a part of this event, the Office of Nuclear Reactor Regulation (NRR)
granted temporary relief so that the licensee did not have to perform the
radiographic test of the repair weld as required by the ASME Code.
The
performance of the test would have required a plant shutdown.
The relief
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was based on the satisfactory ic.agnetic particle examination of each weld
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layer and an ultrasonic test of the completed repair which verified that
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the pipe wall thickness was greater than the minimum required.
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6.
10 CFR Part 21 Report Followup
a.
The NRC inspector, by review of documents and discussions with
licensee personnel, verified that the'10 CFR Part 21 reports
discussed below had been reviewed and appropriately acted on by the
licensee.
(Closed) P21-1987-87-19:
Design Defect in Limitorque Valve Operators
Manufactured Prior to 1975
The licensee determined that the limitorque operators at Wolf Creek
Generating Station (WCGS) were manufactured after 1975 and that they
have a machined notch in the belleville spring assembly to prevent
the hydraulic locking problem reported in the 10 CFR Part 21 report.
This information was documented in WCGS Industry Technical
Information Review and Evaluation (ITIP) No. 00360.
(Closed) P21-1987-87-06:
Stationary Sleeve on MSIV Thrust Bearing
Eg ending Past Rotating Face
This 10 CFR Part 21 report reported a defect in reverse seating check
valves manufactured by Atwood and Morrill Company, Inc. and used as
main steam isolation valves (MSIV).
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hydraulically activated, bidirectional, double disc gate valves
manufactured by a different vendor, therefore, this 10 CFR Part 21 is
not applicable to WCGS.
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b.
The NRC resident inspector provided a copy of the 10 CFR Part 21
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report listed below to the licensee for review and action, if
required.
P21-1987-87-65: Wolfe & Swickard-Inner Bearing Race Missing on OG
Air Start Motors
7.
Physical Security Verification
The NRC inspectors verified that the facility physical security plan (PSP)
is being complied with by direct observation of licensee facilities and
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security personnel.
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The NRC inspectors by observation of randomly selected activities verified
that search equipment is operable, that the protected area barriers and
vital area barriers are well maintained, that access control procedures
are followed and that appropriate compensatory measures are followed when
equipment is inoperable.
No violations or deviations were identified.
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8.
Radiological Protection
By performing the following activities, the NRC inspectors verified that
radiologically related activities were controlled in accordance with the
. licensee's procedures and regulatory requirements:
o-
Reviewed documents such as active radiation work permits and the
health physics shift turnover log.
o
Observed personnel activities in the radiologically controlled
area (RCA) such as:
Use of the required dosimetry equipment,
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" Frisking out" of the RCA, and
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Wearing of appropriate anti-contamination clothing where
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required.
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Inspected postings of radiation and contaminated areas.
o
Discussed activities with radiation workers and health physics
supervisors.
Selected NRC inspector observations are discussed below:
On August 28, 1984, the licensee received from Science Applications
International (SAI) a boron analyzer which had a fission chamber in it
that contained approximately 2 grams of Uranium 235 (U235).
This analyzer
was part of the post-accident sampling system.
At the time the analyzer
was shipped and received the fission chamber did not get classified as
special nuclear material (SNM) and therefore the transaction was not
documented in accordance with the instructions in NUREG/BR-0006,
Revision 2, " Instructions for Completing Nuclear Material Transaction
Reports." NUREG/BR-0006 requires that the shipper, in this case SAI,
initiate a DOE /NRC Form 741 and send it to the receiver, in this case the
licensee, with the SNM.
The receiver then fills out the appropriate
section of the DOE /NRC Form 741 and distributes it per the instructions in
Recently, licensee personnel questioned whether or not the
fission chamber should have been treated as SNM and in subsequent
discussions with SAI, it was decided that it should have been.
SAI on
August 19, 1987, sent the licensee a DOE /NRC Form 741 for the fission
chamber which the licensee filled out and distributed in accordance with
No violations or deviations were identified.
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Engineered Safety Features (ESF) System Walkdown
The NRC inspectors verified the operability of ESF systems by walking down
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selected accessible portions of the systems.
The NRC inspector verified
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valves and electrical circuit breakers were in'the required position,
power was available, and valves were locked where required.
The tJRC
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inspectors also inspected system components for damage or other conditions
that could degrade system performance.
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The ESF system walked down during this inspection period and the documents
utilized by the NRC inspectors during the walkdown are listed below:
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System
Documents
Component Cooling Water (EG)
Drawing M-12EG01(Q), Revision 2,
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" Piping and Instrumentation Diagram
Component Cooling Water System"
Drawing M-12EG02(Q), Revision 1,
" Piping and Instrumentation Diagram
Component Cooling Water System"
Drawing M-02EG03(Q), Revision 17,
" Piping and Instrumentation Diagram
Component Cooling Water System"
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Checklist (CKL) EG-120, Revision 8,
" Component Cooling Water System Valve,
Switch, and Breaker Lineup"
SYS EG-120, Revision 5, " Component
Cooling Water System Startup"
SYS EG-203, Revision 5, " Chemical
Addition to Component Cooling Water
System"
SYS EG-201, Revision 4,." Transferring
Supply of CCW Service Loop and CCW
Train Shutdown"
STS EG-001, Revision 4, " Component
Cooling Water Valve Check"
STS EG-100A, Revision 2, " Component
Cooling Water Pumps A/C Inservice Pump
Test"
STS EG-1008, Revision 2, " Component
Cooling Water Pumps B/D Inservice Pump
Test"
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Selected NRC inspector observations are discussed below:
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Three 4-inch drains with flanges on each train of CCW were not
included in CKL EG-120.
o
Drain Valves EG-V381, 382, and 384 were not in CKL EG-120.
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Flow Indicator EG F1-64 was not reading zero when that portion of the
system was isolated.
The observations noted above were given to the licensee.
No violations or deviations were identified.
10. Allegation Followup
(Closed) Allegation (4-87-A-065): An alleger stated that when the plant
was under construction, stainless steel welds were quenched with water.
Findings: The NRC inspector interviewed licensee engineers and determined
that in May 1983, Quality Surveillance Report SR-437M was initiated to
document craft personnel applying a saturated cloth to a stainless field
weld. Although this was standard recommended practice and did not violate
code and/or procedural criteria, it was not being done in accordance with
an approved procedure. The licensee issued guidance for the welders to
use in quenching; however, the licensee subsequently decided to
discontinue quenching.
Conclusion:
The allegation that Wolf C:eek construction quenched welds
was substantiated; however, quenching the stainless steel welds had no
impact on safety.
11.
Followup on Regional Requests for Information
During this inspection period, Region IV management informed the NRC
inspectors of a containment temperature problem at another nuclear power
plant.
It appeared that for a number of years the ambient air temperature
inside containment had been excessively high and the licensee at that
facility failed to adequately address the causes of the high temperatures,
their affect on equipment qualification (EQ)e and on the accident
analysis.
WCGS TS 3.6.1.5 requires that the primary containment temperature be
determined by averaging the temperatures of the four containment coolers
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inletsandthatitnotbegreaterthanf20F. During this inspection
period, the average temperature was 101 F.
When questioned further by the
NRC ;nspector, licensee personnel stated that the temperatures of
individual compartments inside containment could not be determined;
however, Plant Modification Request (PMR) 1975 had been written to place
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some temporary' temperature probes insideLcontainment to monitor'various
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locations during the next cycle.
This item will remain open pending NRCs.
review of the next cycle temperature data (482/8720-03).
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Open Items
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Open items are. matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which involve some action
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on the part of the NRC or licensee or.both.- One open item disclosed
Lduring the inspection'is discussed in paragraph 11.
13.
linresolved Items
Unresolved items are matters about which more information is required in
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order to ascertain whether they are acceptable items, items of
noncompliance, or deviations.
One unresolved item disclosed during the
inspection is discussed in paragraph 5.
14. -Exit Meeting
The NRC' inspectors met with licensee personnel to discuss'the scope and
findingstof-this inspection on' September 4, 1987.
The NRC inspectors also
attended-entrance / exit meetings of.the NRC region-based inspectors
identified below:
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Inspection
Area-
Inspection
Period-
Inspector
Inspected
Report No.
8/3-7/87
R. Stewart
Temporary
87-18
Instruction
Followup.
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