ML17292B225
| ML17292B225 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292B221 | List: |
| References | |
| 50-397-97-18, NUDOCS 9801220170 | |
| Download: ML17292B225 (32) | |
See also: IR 05000397/1997018
Text
ENCLOSU
E
U.S. NUCLEAR REGULATORYCOMMISSION
REGION IV
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location
Dates:
Inspectors
Approved By
50-397
'I
h
- 50-397/97-18
'Washington Public Power Supply System
~
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I
Washington Nuclear Project-2
Richland, Washington
.: ,'ovember 9 through December 20, 1997
If
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'. A. Boynton*, Seriior Resident Inspector
G. W. Johnston, Senior Project Engineer
T. O. McKemon, Senior Operator License Examiner
G. M. Good, Senior Emergency Preparedness
Inspector
H. J. Wong, Chief, Reactor Projects Branch E
Attachment:
Supplemental Information
980i220i70 '980i i5
ADQCK 05000397
8
EXEC TIVE S
MMARY
Washington Nuclear Project-2
NRC Inspection Report 50-397/97-1 8
~Oer
ions
A number of inspectoridentified deficiencies in the control of transient equipment
indicated weak implementation of the licensee's program to prevent seismic interactions
between the equipment and safety-related components.
Three examples of a violation of
plant procedures were identified (Se'ction'G1;1):-
ai
The licensee's material condition inspection program was not fullyimplemented to
maintain and assess
those areas of the reactor building not routinely accessed
by plant
personnel.
As a'result, a lower standard was established for these areas and equipment
and housekeeping
deficiencies were allowed to persist (Section M2.1).
The methodology utilized by the licensee for testing the control room emergency
.
charcoal filters was identified as being from a different, more recent version of the
standard specified in Technical Specifications (TS). Based, in part, upon the staffs
acceptance of the version of the standard utilized by the licensee, and the more
conservative results produced by its methodology, the noncompliance was viewed as a
minor violation (Section M8.2).
incjiiny~rin
~
Identified performance issues in the leakage surveillance and prevention program
regarding plant staff knowledge, program implementation, and procedural
inconsistencies,
were indicative of weak management
involvement (Section E1.1).
The licensee's use of an uncontrolled database
during its power uprate implementation
resulted in an affected design calculation for the ultimate heat sink being missed in the
, review process.
The existing revision of the calculation bounded the parameters of the
power uprate (Section E8.2).
Pl
u
o
~
Weaknesses
in the licensee's program for monitoring and control of combustibles in the
plant resulted in: (1) materials accumulating in limited access areas without being
properly evaluated or tracked, and (2) reducing the value of the licensee's combustible
loading calculation as a tool in supporting plant modifications due inconsistencies
in
calculation, coupled with a relatively large backlog of modifications to the current revision
of the calculation. A violation of transient combustible material control was identified
(Section F1.1).
'-2-
The licensee's failure to test the control room facsimile machine contributed to an
inoperable piece of emergency response equipment going undetected
until it was
required to be used during an actual event. A noncited violation was identified
(Section P2.1).
'I
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umma
f Plan
u
The plant began the inspection period at 100 percent power. Power was reduced to 75 percent
on December 5 to allow for exchange of deep and shallow control rods and for recovery of a
single control rod whose associated scram accumulator had been replaced.
The plant was
returned to full power on December 6 and remained at full power for the balance of the
inspection period.
01
Conduct of Operations
01.1
Po
en
a.
I
onSc
e 77
During routine plant tours over the months of November and December and during tours
of limited access areas (Section M2.1), a number of equipment storage concerns were
noted. The inspector reviewed the licensee's program for controlling transient equipment
in the plant.
sa
The licensee's program for controlling transient or portable equipment in the plant is
described in Plant Procedure Manual (PPM) 10.2.53, "Seismic Requirements For
Scaffolding, Ladders, Man-Lifts, Tool Gang Boxes, Hoists, and Metal Storage Cabinets,"
Revision 16. PPM 10.2.53 provides specific instructions for the proper storage of the
subject items to prevent them from damaging safety-related equipment in a seismic
event.
For equipment covered under PPM 10.2.53, the following storage concerns were noted:
~
Three ladders in the reactor water cleanup (RWCU) heat exchanger room were
found extended and staged for use.
Tw'o additional ladders in the RWCU pump
room mezzanine were found extended and staged for use.
Upon subsequent
entries by licensee personnel, the ladders were properly laid down on the floor.
Based upon the specific location of the ladders, the possibility of damage to
safety-related equipment was considered to be low. Section 7.2.1 of
PPM 10.2.53 requires ladders to be stored in an assigned storage location when
no longer needed for a job.
In addition, ifa ladder does not have an assigned
storage rack, PPM 10.2.53 provides that the ladder shall be laid down on the
floor. The as-found condition of the ladders in the RWCU equipment rooms did
not conform to the requirements of PPM 10.2.53 and is the first example of a
violation of TS 5.4.1 (VIO 50-397/97018-01) for the failure to follow procedures.
0
0
-2-
Scaffold equipment carts were found unsecured
in the Division I emergency
diesel generator room and the diesel building corridor. Specifically, the carts',
brakes were not adequately engaged to prevent the carts from rolling and
potentially impacting safety-related equipment nearby.
From discussions with the
personnel using the carts, it was determined that attempts were made to engage
the brakes; however, the brakes were ineffective. Section 7.2.2.d of PPM 10.2.53
,requires these types of carts to have at least two wheels immobilized to prevent
them from rolling during a seismic event.
The failure to adequately-secure
the
scaffold carts is the second example of a violation.of TS 5.4.1'(VIO. 50-
397/9701 8-01).. >,
.
c;;
.
A large (approximately 7-foot high) storage cabinet was found on the 501 foot
elevation of the reactor building, in close proximity to safety-related equipment.
The cabinet was unsecured and had the potential, ifoverturned, to impact and
damage several instrument lines associated with the reactor recirculation system
and leakage detection system.
The cabinet had been in place for several years.
Section 7.2.2.a of PPM 10.2.53 requires cabinets of this size to be secured or
stored greater than or equal to the cabinet's full height plus 12 inches from any
safety-related equipment.
The failure to properly secure or place the storage
cabinet was identified as the third example of a violation of TS 5.4.1
(VIO 50-397/97018-01).
Several additional examples were identified by the inspector in areas which were
covered by PPM 10.2.53, but which did.not contain safety-related equipment.
Specifically, a scaffold cart and a large portable electrical load bank were found
unsecured
in the corridor outside the vital electrical switchgear rooms on the 467 foot
elevation of the radwaste building. A large gang box was also found in the storage area
adjacent to the spent fuel pool that was in close proximity to spent fuel pool cooling flow
instrumentation sensing lines. Although these conditions did not conform to the
requirements of PPM 10.2.53 and were indicative of weak licensee performance in this
area, they were not considered violations of NRC requirements.
~on I~ion.>
A number of identified deficiencies in the control of transient equipment indicated weak
implementation of the licensee's program to prevent seismic interactions between the
equipment and safety-related components.
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a.
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ci
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The inspector observed the monthly Diesel Generator
1 operability test performed in
accordance with Procedure OCP-Elec-M701, "Diesel Generator
1 Monthly Operability
Test," Revision 2.
bs
ai
san
Fin
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s
/
/
During performance of the monthly operability test, the inspector observed good
coordination of the test between the control room operator and the equipment operator
station at the diesel generator local control station. The reactor operator maintained
communications with the in-plant operator, exercised good three-way communications,
practiced good self-checking practices, and utilized the control room supervisor as a peer
checker to verify correct hand switch manipulations.
The control room operator started,
paralleled, and loaded the diesel generator successfully and ran the diesel generator for
the prescribed
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> duration at full load. Other technicians took the required lube oil
samples and vibration reading. Afterthe
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, the diesel was unloaded and
power to the 4160V bus was transferred back to normal supply.
No test anomalies were
identified during the test.
C.
QQQQJQ~IO
The monthly Diesel Generator
1 operability test was run successfully.
Control room
operators demonstrated
good communications and performed good self-checking and
peer checking.
02
Operational Status of Facilities and Equipment
02.1
n in e ed
fe
Fe
eS
e
Walk
wn
71
The inspectors walked down accessible portions of the following safety systems:
Reactor Core Isolation Cooling
Residual Heat Removal, Trains A and B
Containment Atmosphere Control, Trains A and B
The systems were found to be in the appropriate configuration for the current plant
conditions.
Material condition of the systems was generally good with two exceptions:
Valves RCIC-V-64 and RHR-V-42Awere found with active packing leakage that had not
been previously identified. These are discussed further in Section M2.1.
e
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08
Miscellaneous Operations Issues (92901)
08.1
lo
Viol
'
/9
-: failure to adhere to procedures related to log
entries and record retention.
The inspector verified that the licensee had taken the
applicable corrective actions committed to,in a letter to the NRC dated
November 1, 1996, and Notice of Violation response letter dated October 14, 1996.
R
08.2
osed
Viola i
-
9
1 - 3: failure to adhere to procedures for reactor startup.
This violation involved the failure to followprocedures during an approach to criticality on
July 27, 1996,, when both a shift turnover and shift.brief were conducted during the
approach to criticality.
During this inspection, the inspector verified that the applicable corrective actions
committed to by the licensee in a letter to the NRC dated November 1, 1996, had been
completed.
083
Additionally, the inspector verified other commitments, which had not yet been fully.
completed, were being tracked and scheduled for completion.
For example, the licensee
was in the process of developing an analytical tool for estimating critical position based
upon subcritical multiplication. This action was scheduled for completion by
January 1,1998.
e
iola i
-
6 -: failure to have an appropriate procedure for
adjustable speed drive testing to preclude a nonlicensed person from operating the
reactor controls. The inspector verified that corrective actions that the licensee
committed to in Notice of Violation response letter dated December 6, 1996, and in a
letter to the NRC dated November 1, 1996 (Items 5, 18, 22, 23, and 24), had been
completed.
08.4
ose
ER
- inadvertent main steam isolation valve (MSIV)closure
while in cold shutdown.
The licensee determined that the root cause of the closure of the
MSIVs involved deficiencies in the governing surveillance procedure that was testing the
MSIVisolation logic to the reactor protection system.
Poor communications between the
two operating crews involved with the testing and the lack of a questioning attitude when
faced with unexpected results were also considered to be contributing factors.
In response to the event, the licensee verified that the closure signal was an artifact of
the test procedure and not required by actual plant conditions. The procedure was
subsequently revised and the testing was completed satisfactorily.'he procedure
preparation process was also revised to require verification that a procedure adequately
addresses
each plant operating condition in which the procedure may be performed.
To
address the contributing factors, operators were counseled on the expectation for a
questioning attitude when faced with unexpected plant response and to involve other
crew members when dealing with unexpected alarms.
Recognizing that the first
-'5-
operating crew involved with the testing.was aware, of the procedural deficiencies,
expectations were also reemphasized for the need to promptly correct-procedural
weaknesses.
a
n
M1.1
Conduct of Maintenance
'
I<
6
7612
I
The inspectors observed and/or reviewed the followingwork activities:
Drywell Pressure
Instrument (CMS-PT-8) Replacement (Work Order ¹DXG3)
OSP-ELEC-M703, High Pressure Core Spray Diesel Generator Monthly
Operability Test
Repack of Valve MS-V-20 (Work Order ¹JKR8)
Repair of High Pressure Turbine Flange Leak (Work Order ¹JKZ6)
Good coordination was noted between the disciplines for each of the work activities.
Strong engineering and health physics involvement was noted in the repack of the main
steam valve and the repair of the high pressure turbine leak. Coordination of these
activities with the reactor downpower on December 5 provided for reduced area dose
rates and a dose savings.
Personnel qualifications were verified and the appropriate TS action statement was
entered for replacement of the drywell pressure transmitter and for the repacking of
Valve MS-V-20.
M2
Nlaintenance and Material Condition of Facilities and Equipment
M2.1
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Tours were conducted of specific areas of the reactor building and turbine building that
are infrequently accessed
during power operations.
The areas included normally locked
-'6-
high radiation and/or contamination areas.
Emphasis was placed upon examining
material condition of equipment and evaluating the impact of transient equipment and
material.
cl 'l'
Deficiencies were identified to varying'degrees
in each of the areas inspected.
Taken
individually, a majority of the deficiencies were considered minor with minimal or
negligible impact on equipment reliability. However, several of the deficiencies were
considered to be more than minor and, taken collectively, the number of deficiencies
indicated an overall weakness
in the licensee's maintenance-and
housekeeping of these
areas.
I
I
The following material condition deficiencies were identified:
Active packing leakage was identified on Valve RCIC-V-64. Valve RCIC-V-64
originally provided for steam admission to the now defunct steam condensing
mode of the residual heat removal (RHR) system.
The valve is deactivated in the
closed position and performs the safety function of containment isolation. Further
discussion of this valve and the licensee's programs for evaluating its condition
are provided in Section E1.1
Minorflange leakage (one drop per minute) was identified on the'upstream
piping
of the RWCU nonregenerative
heat exchanger.
Packing leakage was also
identified on Valves RWCU-VR and RWCU-V-704. Valve RWCU-V-4 is the
outboard containment isolation valve to the RWCU system, while Valve RWCU-
V-704 is a pressure gage root valve downstream of Valve VC.
Operational
leakage from this system is undesirable due to the relatively high activity levels in
process fluid.
The inspector noted that packing leakage from Valve RWCU-V-704 contributed to
elevated reactor building effluent levels early in the current operating cycle. As a
result, the licensee made an entry into the RWCU pump rooms to tighten the
valve's packing in August 1997.
Valve RHR-V-24A (Suppression
Pool Cooling/Test Return), RHR-V-27B
(Suppression
Pool Spray), and RHR-V-42A (Low Pressure Coolant Injection
Isolation) each showed signs of packing leakage.
Active leakage (10 drops per
minute) was observed on Valve VQ2A, while standing water was observed
around the packing and bonnet wells of Valves V-24A and V-27B. Leakage from
these components is undesirable as each of the valves is within the boundary of
the postloss-of-coolant accident RHR recirculation loop.
A number of manual Borg-Warner valves were identified without their t-handle
operator installed, including Valves RHR-V-708A and C. Although no reference
0
was identified within the emergency operating procedures for manipulating these
valves, the lack of a'manual operator unnecessarily.complicates
system
alignment and alignment verification.
Lighting deficiencies, were identified in each'of the areas inspected, the'most
significant being the RHR Loops A and.C pipe chase on the reactor building
522 foot elevation where none of the permanently installed lighting was
functional. Occasional access to this area is required for emergency core cooling
system venting and RHR system leakage inspection.
The following housekeeping
deficiencies, were observed:
~
Five ladders were identified as being improperly stored in areas containing
safety-related equipment.
The improper storage of the ladders was a violation of
licensee procedures and is discussed
in Section 01.2.
~
Parts from an abandoned
pipe support, constructed of heavy-gage steel plate
and 4-inch x 4-inch steel box beam, were discovered on a mezzanine in'the
southwest valve room on the 471 foot elevation of the reactor building. A review
of plant records found that the pipe support was abandoned
in 1986. The
improper storage of this material was considered significant in that, during a
seismic event, the potential existed for the material to fall from the mezzanine on
to some small-bore RHR piping below.
In the traversing incore probe mezzanine and the 492 foot RHR horizontal pipe
chase, quantities of transient combustibles were discovered in excess of the
allowable limits established by licensee procedures.
Transient combustibles
were also discovered in several other areas, including the RWCU heat exchanger
room and the RWCU pump room mezzanine.
However, a thorough inventory
was not completed to determine whether or not those quantities were within
limits. Further discussion of transient combustibles and the licensee's fire
protection program are provided in Section F1.1.
An empty beer can was found in the 492 foot RHR horizontal pipe chase and
cigarette butts were found in the southwest valve room on the 471'levation of
the reactor building. Information obtained from the bottom of the can showed that
the beer was produced in 1995. The age of the cigarette butts was unknown.
Although the specific circumstances surrounding these items cannot be
determined, their presence
is of concern in that they are prohibited to be
consumed inside the radiologically controlled area.
Furthermore, licensee
procedures prohibit alcoholic beverages
on the plant site.
Based upon the number of discrepancies
identified, the inspector reviewed the licensee's
program for evaluating plant material condition. This program is described in
PPM 1.3.19, "Plant Material Condition Inspection Program," Revision 21. The plant
general manager has overall responsibility for this program.
PPM 1.3.19 provides
0
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guidelines and examples for inspecting material condition, housekeeping,
radiological
conditions, and fire protection.
Department:managers
are informally assigned specific
areas of the plant by the plant manager and are expected to include out-of-the-way and
limited access areas as part of their assigned inspection. The program, however, does
not provide specific guidelines on the frequency in which inspections are to be performed
and no information was found to indicate that department managers had performed
recent inspections in the areas'walked down b'y the inspector.
Had inspections been
performed in these areas, the guidance in PPM 1.3.19 was found to be sufficient to
identify each'of the above discrepancies."
For high radiation ar'eas', PPIVI'-1.3.19 provides the radiation protection manager the
discretion in determining the'frequency of inspection..'owever,
from discussions with
the radiation. protection manager, it was found'that'he had not been familiarized with the
program or his'responsibilities.
Plant procedures would require specific radiation work
permits to be developed for'these inspections; however, none could be found for
calendar year 1997.
From disc'ussions with several plant staff, including health physics, engineering, and the
plant general manager, it appeared that conditions went unidentified or uncorrected due
to a perceived difference in expectations for compliance with housekeeping
procedures
for those areas not routinely accessed
due to radiation exposure concerns.
It was noted
that operations, maintenance,
and engineering personnel had made entries into several
of the areas for emergent maintenance during the current cycle; however, no material
condition or housekeeping
deficiencies were documented during those entries.
The lack
of management
inspection in these areas was considered to be a contributor to the
difference in standards.
The licensee's material condition program was not fullyimplemented to maintain and
assess
those areas of the reactor building not routinely accessed
by plant personnel.
As
a result, a de facto lower standard was established for these areas, and equipment and
housekeeping
deficiencies were allowed to persist.
Miscellaneous Maintenance Issues (92902)
I
V'
00 -: failure to preplan, document, brief, and authorize
troubleshooting.
This violation involved troubleshooting work outside the defined scope
of the plan. As a result of a human error during the work, the plant experienced a half-
scram actuation.
During this inspection, the inspector verified that licensee corrective actions committed to
in the licensee's response to the NRC in a letter dated May 1, 1996, had been
completed.
The incident had been reviewed with involved individuals, instrumentation
and control technicians had been counseled on self-checking practices, and applicable
procedures had been revised.
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M8.2
I
e
s ecti
liow
em
I
50-39
00 -: the licensee's TS 4.6.5.3.b.2
directs testing of engineered safety feature (ESF) charcoal filters in accordance with
ANSI N509-1976; which requires, in part, the filter bed to be heated to 80'C. However,
the licensee's Final Safety Analysis Report (FSAR), Section 6.5, indicates that charcoal
filtertesting would be accomplished using ANSI N510-1980, which requires, in part, that
the filterbed be heated to 30'C.
As part ofthe licensee's conversion to Improved Technical Specifications (ITS),
implemented in'March 1997, the specific testing requirements for the ESF filters were
relocated to TS 5.5.7, "Ventilation Filter Testing Program." TS 5.5.7 now references
ANSI N510-1989 as the applicable version of the testing standard.
Th'e 1989 standard
did not change fhe te'sting methodology from th'e 1980 standard.
The staffs approval of
the Ventilation Filter Testing Program in ITS indicated that the testing methodology of the
later versions of the ANSI standa'rd would provide confidence that the filters would
perform their safety function. Additionally, it was found that testing of the charcoal beds
at lower temperatures
provided more conservative results in that, at lower temperatures,
the activated charcoal has a lower affinityfor iodine retention and, thus, a higher leakage
rate through the bed.
Based upon the staffs approval ofthe licensee's ITS and its use of ANSI N510-1989 and
the fact that the more recent versions of the standard, utilized by the licensee, provide a
methodology that produces more conservative results, the failure to test ESF charcoal
filters using the 1976 version of the ANSI standard was not considered to be safety
signiTicant. The licensee's actions to implement ITS effectively addressed
the
discrepancy between TS and the FSAR. This failure constitutes a'violation of minor
significance and is being treated as a noncited violation, consistent with Section IVof the
NRC Enforcement Policy (NCV 50-397/97018-02).
II. E
ineer
E1
Conduct of Engineering
E1.1
ool
Lea
1 7
In response to packing leakage identified on Valve RCIC-V-64 (Section M2.1), the
inspector reviewed the licensee's program for controlling primary coolant leakage outside
containment.
b.
aio
andFi din s
The requirement to evaluate and reduce leakage'from systems outside containment
stems from TMIAction Item III.D.1.1, as described in NUREG-0737. The licensee
addressed
this item in Appendix B of their FSAR and incorporated the requirements in
-10-
the TS. TS 5.5.2, "Primary Coolant Sources Outside Containment," outlines the
requirements of the prograrri, including preventive maintenance,
visual inspections, and.
integrated leak tests.
PPM 1;5.6, "Leakage Surveillance and Prevention Program," Revision 8, implements'the
requirements of TS 5.5.2. 'Ih reviewing PPM 1.5.6, the inspector noted a,number of
discrepancies
between the procedure, TS, and the FSAR.
It was also noted that
Valve RCIC-V-64 (discussed
in Section-M2.1) was not covered by PPM 1.5.6. Some of
the specific discrepancies
identified include:
N
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PPM 1.5.6 directs visual inspections of the applicable systems during routine
plant operations, including surveillances.
Visual inspection requirements were
verified to be included in. system surveillance procedures,
but were not included
in system operating procedures.
The visual inspection requirements in
TS 5.5.2 and the FSAR were not as clearly defined and, thus, subject to some
interpretation.
Although PPM 1.5.6 indicates the need to inspect systems during
routine operations, the licensee contended that visual inspections were only
intended to be performed during surveillances.
That position does not conflict
with TS requirements or the language in the FSAR.
~
Visual inspections of selected instrument racks are required by PPM 1.5.6 during
operator rounds.
However, the specific instrument racks required to be inspected
are not defined by the procedure, nor are they captured in operating instructions
for equipment operators.
During.routine plant tours, the inspector did not identify
any leakage from safety-related instrument racks.
The lack of specific guidance
on inspecting instrument racks was considered a program weakness
in
PPM 1.5.6.
TS 5.5.2 speciTically references the standby gas treatment system as one of the
systems covered under the leakage surveillance program. As such, periodic
maintenance,
visual inspections, and integrated leak tests are required.
PPM 1.5.6 also references the standby gas treatment system; however, no
requirements are included for performing visual inspections and integrated leak
tests.
At the end of the inspection, insufficient information was available to
reconcile this discrepancy and, therefore, an unresolved item was opened
(URI 50-397/97018-03).
TS 5.5.2 requires integrated leak tests to be performed on a 24-month interval.
The leakage surveillances are listed as biennial procedures in Attachment 6.1 to
PPM 1.5.6.
However, Section 3.2.1 and Note 2 of Attachment 6.1 both state that
the surveillances are performed on an 18-month cycle. From discussions with
the program manager, it was determined that PPM 1.5.6 was not adequately
updated upon the conversion to ITS.
A review of the most recent integrated leakage surveillances for the systems covered by
TS 5.5.2 showed that the systems were relatively leak-tight. That conclusion was
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supported by a low number of leakage-related work requests against those systems.
As
noted in Section M2.1; packing leakage was identified for several valves located in high
radiation/contamination areas.
Valves RCIC-V-64 and RHR-V-42A both had identifiable
.
packing leakage that had not been previously documented by the licensee.
Allof the integrated leakage surveillances were found to be current, with the exception of
the surveillance for reactor core isolation cooling'(RCIC). The latest RCIC surveillance
was in the review process, while the previous surveillance had been completed in June
1995.
Based upon the original 18-month frequency required by TS 6.8.4.a.2, the RCIC
surveillance was overdue in. January 1997.. The surveillance became overdue based
upon the licensee's surveillance scheduling program applying a 25 percent extension to
the surveillance interval (application of TS 4.0.2).
However, TS 6.8.4.a.2 provides no
basis for allowing the use of TS 4.0.2 in scheduling the surveillances.
The practice of
applying an extension period to these surveillances had apparently been in place since
initial facility licensing. Atthe end of the inspection period, the licensee was reviewing
other TS administrative programs to determine ifTS 4;0.2 had been inappropriately
applied in other situations.
Pending the licensee's further review of other TS programs,
this is an unresolved item (URI 50-397/97018-04).
The inspector noted that, through approval of the licensee's
ITS, implemented in March
1997, the staff authorized the extension of the surveillance interval to 24 months and
allowed the use of Surveillance Requirement 3.0.2 to extend the interval by 25 percent
for scheduling flexibility. Thus, the licensee has already taken action to 'explicitly
authorize the historical scheduling practice for the leakage surveillance and prevention
program.
In discussing the above issues with licensee personnel, including managers,
knowledge
gaps were identified with respect to the current status of the program and its
implementation.
Specifically, the program manager was unaware of the discrepancies
noted in the procedure and was not knowledgeable of the outstanding leakage-related
work requests for those systems covered under the program. The operations manager
was unfamiliar with his responsibility to ensure periodic visual inspections are performed
for the systems included in the program, while the system engineering manager,
responsible for overall program implementation, was unfamiliar with the requirements of
PPM 1.5.6.
QggcIugiggs
Performance issues regarding plant staff knowledge, program implementation, and
procedural inconsistencies were indicative of weak management involvement and poor
maintenance of the primary coolant system leakage surveillance and prevention
program.
Miscellaneous Engineering Issues (92903)
-12-
E8.1
E8.2
Clo e
IFI
-3 7
00 -0: failure to'update the FSAR when the standby service
water, keep-full system was abandoned
in place.
This, issue was converted to an
unresolved item in NRC inspection Report 50-397/96-201 and willbe dispositioned
through that report item. This IFI is administratively closed.
s
- .the licensee had not performed a review of.
Calculation ME-02-85-75, Revision 2, which analyzes the heat removal capacity of a
single service water pond during an accident, prior to implementing the reactor power
uprate during the spring 1995 refueling outage.
I'
The, licensee determined the cause of the failure to perform a review of the calculation
was the incorrect title entry in the database that was utilized to track reviews. The
abbreviation for the. service water system utilized by the Calculation Review was "SSW."
The database
used during the review for the power, uprate identified this acronym as
"Sacrificial Shield Wall." The incorrect title in the database
led the reviewers, relying on
the database,
to erroneously assume that the identified calculation was unaffected by the
The licensee conducted a further review of the database,
concluding that
there were no other cases where a system was erroneously identified. The database
used in this instance was a working file on a reviewer's computer and not part of one of
the licensee's sitewide databases.
Further, the database was not formally controlled by
the licensee.
The inspector noted that the reliance on this database
appears to be a
weakness
in the licensee's performance of the power uprate review.
The licensee noted in Problem Evaluation Request (PER) 296-0787, that the service
water pond calculation, ME-02-85-75, Revision 2, used a conservative heat generation
rate that bounded the power uprate conditions.
The licensee further indicated that ME-
02-85-75 will be revised to reflect the current conditions, which include the power uprate.
The licensee's corrective actions were considered appropriate.
The erroneous entry of
SSW in a database
appeared to have been an isolated occurrence.
The inspector
concluded the licensee's review of affected calculations was weak in that it relied upon
an uncontrolled database.
E8.3
sed
-397/
33-: the licensee's evaluation of the standby service
water (SSW) system leaks, the quantifying and evaluation of hydraulic transient (water
hammer) loads of the SSW system, and the control of SSW flow control valve opening
times were examined.
On November 29, 1995, an NRC inspector and a licensee equipment operator identified
a leak from the weld area of a 3/4-inch Sock-o-let fitting, near the high point of the
service water line Loop B. A stream of water was spraying from the leak onto an
adjacent motor control center.
The licensee initiated PER 295-1229 to resolve the issue.
The evaluation performed for the leak indicated that the leak occurred due to wet/dry
cyclic conditions in the crevice area of the socket weld. In addition, the local crevice
corrosion rate was affected by microbiological-induced corrosion. The causes of the leak
were attributed to inadequate corrosion monitoring and chemical treatment for corrosion
-'1 3-
and biological control. Further, a contributing cause was the crevice of the socket weld
design of the SSW small bore piping.', The licensee dispositioned the leak by initiating
=
rework and the affected Sock-o-let flittingwas replaced.
The licensee has subsequently
implemented an ongoing program to reduce microbiological-induced corrosion in the
affected SSW and high pressure core spray service water (SW) systems.
The licensee conducted a visual inspection of SSW Loop B following-the troubleshooting
associated with the SSW pinhole leak on November 29, 1995.
Personnel, including
equipment operators, the system engineer, the shift support supervisor, and an NRC
inspector, observed the restart of the'SW line Loop B., The observers heard a bang in
the system, which the inspector characterized to the licensee as an apparent hydraulic
Following the startup, the licensee inspected the SSW
Loop B restraints and supports, finding no apparent damage.
In a continuing effort to evaluate the water hammer, a startup of Loop Awas conducted,
with similar results. The licensee noted a bang in Loop A, but of a lower magnitude than
that of Loop B. Further investigation by the licensee identified a difference in opening
times for the butterfly valves, which control service water flow at the initiation of SW
system startup.
The control circuitry of the butterfly valves limits their opening during the
initial seconds of system startup in order to limitthe initial flowof water and to soften its
impact on the partially voided SW system.
Subsequent testing indicated the valves were
operating within their specified operating times and met the inservice testing
requirements.
Further investigation revealed a temperature control valve to
Chiller CCH-CU-1 B may have been the cause of the water hammer events.
When room
temperature
in the chiller room was less than 70'F, Valve SW-TCV-15B would remain
closed. A change to the setpoint of the valve had occurred previous to the events from
55'F to 70 F. With a space heater in place in the chiller room raising temperature above
70'F, no observation of a water hammer noise was generated.
The licensee has
changed the setpoint of the temperature switch associated with Valve SW-TCV-15B to
63'F. This change was considered appropriate to assure that the valve does not
contribute to future water hammer events.
The inspector began a review of the licensee's analysis of SW Loop B piping. This was
documented in Calculation Modification Record (CMR) ME-02-96-25. The calculation
was to evaluate the SW Loop B piping hydraulic transients in response to
PER 295-1275.
The inspector noted that the results of the calculation did not indicate
there were any substantive stresses
imparted to the piping or supports.
The inspector at
the time of the inspection had not completed a review of the calculational methodology.
This issue willbe tracked as an IFI (IFI 50-397/97018-05).
lo
d
ER
0-397/9
01 -00: the licensee found that the test methodused to satisfy
a RCIC surveillance requirement was incomplete. The TS requirement to verify the
RCIC Division 2 automatic isolation seal-in relay contact function was not covered in the
applicable surveillance procedures.
The RCIC Division 2 functional testing procedures
-14-
were deficient in that none of them verified that the automatic isolation signal would seal
in to maintain the RCIC system inboard steam supply isolation valve closed in the event
of a steam line break.
The licensee immediately performed the appropriate testing upon discovery and revised
PPMs 7.4.3.2.1.80, "RCIC Isolation on RCIC Steam Supply Flow High DIV2-CFT/CC,"
and 8.3.303, "Isolation - Response Time Testing RCIC-RLY-K33, K54, K55, K66 (Valve
Group 8/9) (DIV,2),"to adequately meet the intent of the requirement.
The licensee also
completed a technical adequacy review of all logic system functional tests for seal-in
contacts to ensure that these tests were adequate.
The licensee found that most
surveillance tests used a,test-sequence
that leaves in place the channel trip until
restoration of the circuit logic, which masks the direct.seal-in function verification. All
affected surveillances were corrected.
The inspectors reviewed the corrective actions and found them to be acceptable.
This
nonrepetitive, licensee-identified and corrected violation is being treated as a noncited
violation consistent with Section VII.B.1 of the NRC Enforcement Policy
(NCV 50-397/97018-06).
V
a
F1
Conduct of Fire Protection Activities
F1.1
o
s'be
In response to the identification of uncontrolled transient combustibles in limited access
areas (Section M2.1), the inspector reviewed the licensee's program for controlling
transient combustibles in the plant. The inspector also reviewed the licensee's fire
hazards analysis (FHA) to ensure that the identified combustibles did not exceed that
assumed
in the FHA.
The licensee's fire protection evaluation is described in Appendix B to the FSAR.
Appendix B provides a general overview of the methodology utilized in the FHA and
references Calculation FP-02-85-03 for the specific assumptions utilized in each fire area
for determining combustible loading.
To ensure that transient combustibles in the plant do not exceed that assumed
in the
FHA, the licensee has implemented PPM 1.3.10C, "Control of Transient Combustibles,"
Revision 0. To accomplish this, PPM 1.3.10C provides guidelines for different areas of
the plant to help plant staff determine when transient combustibles need to be evaluated
and controlled through the transient combustible permit process.
Depending upon the
-1 5-
specific fire area, as defined by the FHA, the amount of transient combustibles can be
evaluated against specific guidelines to determine the need for'a permit.
Based upon insights from the FHA, the licensee has designated certain areas of the
plant as "combustible free zones,". and others as "zones of limited combustibles."
For
"combustible free zones," any use of combustibles must be continuously attended and
no storage of combustibles is allowed. For "zones of limited combustibles," stricter
controls are placed upon transient materials.
For most other areas of the power block,
Attachment 9.2 of PPM 1.3.10C defines the amount of material that requires a transient
combustible permit.
P
During tours of limited access areas, the inspector found transient combustibles in
varying amounts in two.areas (the traversing in-core probe room mezzanine and the
492 foot level horizontal pipe chase in the reactor building). In both areas, the amount of
combustibles exceeded the guidelines of PPM 1.3.10C for requiring a transient
combustible permit. However, permits had not been reviewed or issued for either of the
areas.
Based upon the infrequent access to the areas, it appeared that the combustibles
had been in the areas since at least June 1997, during the licensee's last refueling.
outage.
Preliminary indications were that the material was left in the areas based upon
the perception by plant staff that a lower'standard is acceptable
in these areas to support
ALARA(as low as reasonably achievable) considerations.
A review of the FHA showed
that the quantities of material identified in the various areas did not exceed the
assumptions of the FHA and, therefore, the safety significance was considered to be
relatively low. The failure to properly track and control transient combustibles in the two
limited access areas was identified as a violation of TS 5.4.1 (VIO 50-397/97018-07) for
the failure to followprocedures
In reviewing Calculation FP-02-85-03, the inspector identified several discrepancies
in
the assumptions
utilized in the analysis and noted that the calculation was last revised in
1993 with 35 CMRs against it. Although each CMR was evaluated to determine its
impact upon the combustible loading for the applicable fire area, without updating the
base calculation, multiple CMRs affecting a single fire area could lead to an undesirable
cumulative effect that would not be captured in the CMR review process.
At best, the
CMR process becomes cumbersome
in evaluating multiple changes and supporting
plant modifications.
The base calculation also contained several internal discrepancies
and discrepancies
with PPM 1.3.10C.
Specifically, the combustible loading analysis for the Division I 4160V
switchgear room assumes that the silicone-based coolant in the 4160V/480V
transformers is combustible, with a heat release of approximately 60 million Btu.
However, the combustible loading analysis for the Division II 4160V switchgear room
assumes that the same silicone-based fluid is noncombustible and does not contribute to
the heat release.
The analysis for the diesel fuel oil day tank rooms does not provide an
allowance for transient combustible materials bas'ed upon the practice that the rooms are
normally locked.
However, PPM 1.3.10C does not explicitly prohibit transient
combustibles in the day tank rooms and it was noted that the rooms are, in fact, not
-'16-
normally locked. The combustibles found in the day tank rooms, including adsorbents,
wooden dipsticks, and a 2-gallon container of diesel fuel oil, did not exceed the general
limits of PPM 1.3.10C for requiring a transient combustible permit.
In the 480V vital motor control centers, the combustible loading analysis assumed
a
transient combustible loading equivalent to 50 Ibs of paper.
However, PPM 1.3.10C
allows for up to an equivalent of 100 Ibs of paper in this area without requiring a transient
combustible permit.
The review of Appendix B to the FSAR noted'that the specific assumptions and results of
the combustible loading analysis for each fire area were removed from the FSAR in
conjunction with a change to the FHA methodology in,1994.
However, a safety
evaluation, in accordance with 10 CFR 50.59, had not been documented to show that the
change did not constitute an unreviewed safety question.
The licensee initiated a PER to
address the issue and was reviewing the impact of the FSAR changes at the end of the
inspection period. An unresolved item was opened to disposition this issue (URI 50-
397/97018-08).
The licensee's program for controlling transient combustibles in the plant has been
applied inconsistently in that materials have been allowed to accumulate in limited
access areas without being properly evaluated or tracked.
Inconsistencies
in the licensee's combustible loading calculation, coupled with a
relatively large backlog of modiTications to the current revision of the calculation, resulted
in reducing the value of the calculation as a tool in supporting plant modifications.
P2
Status of Emergency Preparedness
Facilities, Equipment, and Resources
P2.1
in
nanc
of
n
27
The inspectors reviewed the licensee's actions to address
PER 297-0242, dated
March 25, 1997,
involving an inoperable control room facsimile machine.
b.
rva ions and Findin
While attempting to make offsite agency notifications during a notification of unusual
event on March 20, 1997, the operator discovered that the control room facsimile
machine was inoperable.
Notifications were made via alternate methods, and the
inoperable facsimile machine was replaced about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> later. The licensee's
investigation disclosed that the last documented preventive maintenance test on the
-17-
facsimile machine was September 1996.
However, PPM 13.14.4, "Emergency
Equipment," Revision 24, Attachment 5.5, requires a monthly check of the control room
facsimile machine.
Based on the investigation results, the licensee took the following corrective actions:
Alltelecommunications service instructions were revised and updated to include
detailed checklists for performing preventive maintenance activities with individual
sign-offs.
~
A 1-year program was being developed to require a telecommunications systems
team lead to observe one preventive maintenance task per month to ensure that
the task was properly performed.
~
A self-assessment
of the telecommunications service procedures and preventive
maintenance program was scheduled to be performed.
~
Procedural compliance training for telecommunications services personnel has
been completed.
~
Procedural compliance training for information services and facilities personnel
has been scheduled.
The failure to perform a monthly check of the control room facsimile machine is a
violation of TS 6.8.1, which requires the licensee to followprocedures that implement the
emergency plan. This nonrepetitive, licensee-identified and corrected violation is being
treated as a noncited violation, consistent with Section VII.B.1 of the NRC Enforcement
Policy (NCV 50-397/97018-09).
~Co
tugign
The licensee's failure to test the control room facsimile machine contributed to a piece of
emergency response equipment being inoperable until it was required to be used during
an actual event. A noncited violation was identified.
V
ana
e
en
Me
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management after the
conclusion of the inspection on December 31, 1997. The licensee acknowledged the findings
presented.
The inspectors asked the licensee whether any materials examined during the inspection should
be considered proprietary.
No proprietary information was identified.
0
AITkCtl
NT
Supplemental Information
PARTIALLIST OF PERSONS CONTACTED
D. Hillyer, Radiation Protection Manager
P. Inserra, Licensing Manager
D. Kobus, Fire Protection, Supervisor
M. Monopoli, Operations Manager
J. Peterson,
Fire Protection Engineer
G. Shindehite, Operations
G. Smith, Plant General Manager
J. Stacks, Engineering
S. Szendre, Raytheon Foreman
INSPECTION PROCEDURES USED
IP 37551:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 82701:
IP 92901:
IP. 92902:
IP 92903:
Onsite Engineering
Surveillance Observations
Maintenance Observations
Plant Operations
Plant Support
Operational Status of the. Emergency Preparedness
Program
Followup - Operations
Followup - Maintenance
Followup - Engineering
0 ene
ITEMS OPENED, CLOSED, AND DISCUSSED
50-397/97018-01
50-397/97018-02
50-397/97018-03
50-397/97018-04
50-397/97018-05
50-397/97018-06
50-397/97018-07
50-397/97018-08
50-397/97018-09
failure to properly secure transient or portable equipment
failure to test control room charcoal filters in accordance with TS
adequacy of suweillance procedures for SGT to meet TS 5.5.2
application of TS 4.0.2 to surveillances required by TS 6.8.4.a.2
IFI
review oftransient hydraulic loads on SSW loop piping
inadequate procedure for verifying RCIC isolation seal-in logic
failure to control transient combustibles
lack of a written safety evaluation for changes made to the FHA
failure to perform required monthly tests of control room facsimile
Qmd
-2-
50-397/95009-01
LER
inadvertent MSIVisolation during testing
50-397/95011-00
LER
50-397/95033-01
IFI
50-397/9600'I-01
50-397/96003-05
.
IFI,
50-397/96006-04
IFI
inadequate procedure for verifying RCIC isolation seal-in logic
review of licensee actions in response to identified SSW system
leakage
failure to pre-plan, document, brief, and authorize
troubleshooting
conformance of.charcoal filtertesting with TS requirements
failure to update the FSAR when SSW keep-full system was
abandoned
50-397/96016-01
failure to adhere to procedures related to log entries and record
retention
50-397/96016-03
failure to adhere to procedures for reactor startup
50-397/96016-05
failure to have an appropriate procedure for adjustable speed drive
testing to preclude a nonlicensed person from operating the
reactor controls
50-397/96024-02
failure to review ultimate heat sink capacity calculation prior to
implementation of reactor power uprate
50-397/9701 8-02
failure to test control room charcoal filters in accordance with TS
50-397/97018-06
inadequate procedure for verifying RCIC isolation seal-in logic
50-397/97018-09
failure to perform required monthly tests of control room facsimile
I
-3-
LIST OF ACRONYMS USED
IFI
, ITS
LER
MSIV,
NRC
PER
'RCIC
TS
WNP-2
calculation modification record
~engineered safety feature
fire hazards analysis
Final Safety Analysis Report
inspection followup item
Improved Technical Specifications
licensee event report
noncited violation
U.S. Nuclear Regulatory Commission
problem evaluation request
Plant Procedure Manual
reactor core isolation cooling
standby service water
service water
Technical Specifications
unresolved item
violation
Washington Nuclear Project-2
~
~
4
5
~