ML17313A391
| ML17313A391 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/22/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17313A389 | List: |
| References | |
| 50-528-98-03, 50-528-98-3, 50-529-98-03, 50-529-98-3, 50-530-98-03, 50-530-98-3, NUDOCS 9806010329 | |
| Download: ML17313A391 (46) | |
See also: IR 05000528/1998003
Text
ENCLOSURE 2
U.S. NUCLEAR REGULATORYCOMMISSION
REGION IV
Docket Nos.:
50-528
50-529
50-530
License Nos.: NPF-41
NPF-'51
Report No.:
50-528/98-03
50-529/98-03
,50-530/98-03
Licensee:
Arizona Public Service Company
Facility:
Palo Verde Nuclear Generating Station, Units 1, 2, and 3
Location:
5951 S. Wintersburg Road
Tonopah, Arizona
March 8 through April 18, 1998
Inspectors:
Jim Moorman, Senior Resident Inspector
Nancy Salgado, Resident Inspector
Dan Carter, Resident Inspector
Frank Brush, Resident Inspector, Callaway
Dave Corporandy, Project Engineer, Branch F
Approved By: Phil H. Harrell, Chief, Projects Branch D
Attachment:
Supplemental Information
'F8060i0329 '&0522
ADGCK 05000528
G
ll
0
E
UTIVE SU
MA Y
Palo Verde Nuclear Generating Station, Units 1, 2, and 3
NRC Inspection Report 50-528/98-03; 50-529/98-03; 50-530/98-03
~O~r~ion
~
Operator oversight and direction of the Unit 1 drain down to midloop evolution, and
decisions to take conservative actions during the evolution, were excellent.
Licensee
activities related to midloop operation demonstrated
a strong safety focus (Section 01.1).
~
Inattention to detail by two licensed operators resulted in improper installation of danger
tags.
This was identified as an isolated example, and as a result, a noncited violation
was identified (Section 01.2).
The licensee demonstrated
good communications and a formal safety approach when
performing refueling operations.
Operations exercised good judgment by suspending
core alterations and movement of irradiated fuel in containment while a containment
integrity issue was being resolved (Section 01.3).
Maintenance of the control room essential filtration system was very good as evidenced
'y the material condition of the components (Section 02.'1).
Unit 2 auxiliary feedwater (AFW) system was installed and maintained per the applicable
design and operating requirements for those portions of the system reviewed.
The
material condition of the AFW system was good (Section 02.2).
Unit 3 containment spray (CS) Train A was installed and maintained per the applicable
design and operating requirements for those portions of the system reviewed.
The
material condition of the CS system was good. The CS system unavailability was very
low (Section 02.3).
Routine and outage-related
maintenance activities were generally conducted in a safety
conscious manner by knowledgeable technicians using approved procedures.
Good
work and foreign material control practices were observed (Section M1.1).
Engineering personnel performed boroscope inspections on Unit 1 Emergency Diesel
Generator (EDG) A cylinder liners without obtaining proper authorization.
This is a
violation of Technical Specification (TS) 6.8.1 for the failure to follow procedure
(Section M1.3).
A violation was identified for the failure to provide specific criteria for licensee personnel
to determine the minimum required level in the pool prior to moving spent fuel. This is a
violation of TS 6.8.1 for the failure to provide an adequate procedure (Section M1.4).
1
0
I
-2-
the permanent repair of a leak in the Unit 1 Steam Generator 2 downcomer sampling line was
consistent with the system design requirements (Section M1.5).
Observable material condition of the three units was satisfactory.
Material condition of
the interior of components disassembled
for the Unit 1 outage was good (Section M2.1).
Material condition of the equipment in the Unit 1 containment was good'(Section M2.2).
EncnE.e~er
The licensee's
10 CFR 50.59 screening and evaluation for the modification to add an
inspection port to Unit 1 Steam Generator
1 was thorough and comprehensive
(Section E1.1).
Licensee plans to address operational and equipment problems resulting from the
inability of some components and computer software to correctly interpret some dates
that occur before and after the year 2000 appeared to be comprehensive
(Section E1.2).
The radiological protection program was effectively implemented in those areas reviewed
(Section R1.1).
Re
o
De
il
Summ
ofPlan Sa us
Unit 1 operated at 100 percent power until March 14, 1998, at which time the unit was shut down
for the planned seventh refueling outage.
The unit was in Mode 3 at the end of this inspection
period.
Units 2 and 3 operated at essentially 100 percent power for the duration of this inspection
period.
01
Conduct of Operations
01.1
'doo /R duced lnvento
Ac'v s
Uni
1
a.
I se 'n
e 71707
On March 17, 1998, the licensee drained the reactor coolant system (RCS) to the
midloop condition using Procedure 40OP-9ZZ16, "RCS Drain Operations," Revision 9.
This was done to allow installation of steam generator nozzle dams in preparation for
eddy current testing of the steam generator tubes.
The inspectors reviewed the
licensee's preparations for midloop operations and observed control room operators as
they performed the various evolutions.
b.
Observations and Find'n
The licensee augmented the onshift operating crew with a team dedicated to perform
=
midloop operations.
The midloop team was comprised of a control room supervisor,
reactor operator, and shift technical advisor, who acted as the midloop coordinator.
There was a clearly defined division of the control room activity oversight between the
midloop team and the normal shift crew. The inspectors'observed
that the midloop team
niaintained positive control of the evolution at all times.
The inspectors reviewed Procedure 40OP-9ZZ16, prior to the reduction of RCS inventory
and verified that all the prerequisites were met.
From discussions with the midloop crew
members, the inspectors determined that the midloop crew had received training and
were knowledgeable of the drain down procedure, contingency plans, and the shutdown
risk assessment.
The inspectors verified that the licensee had calculated and was
sensitive to the short amount of time that existed between loss of shutdown cooling and
boiling in the core.
The licensee minimized unnecessary
work while the unit was in a'reduced inventory
condition. To prevent a loss of shutdown cooling or RCS level perturbations, the
licensee stationed a senior reactor operator at the entry of the auxiliary building to screen
work activities going into the field. In addition, the licensee maintained sources of offsite
and onsite power available, and limited access to critical equipment areas.
I
,(
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During the draindown, at approximately the 102-foot, 8-inch level, the midloop crew
identified discrepancies
in level indication among the locally-installed gage glass and the
refueling water level indication system instruments.
The operators stopped the drain
down and initiated efforts to determine the cause of the level discrepancy.
After the
vented side of the gage glass was purged, the refueling water level indication system
was in agreement with the gage glass within the tolerance allowed. The draindown was
recommenced
and completed successfully.,
Unit 1 entered midloop operation when reactor vessel level dropped below the 103-feet,
1-inch level and was in the midloop condition for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
During that
time, the licensee appropriately implemented and maintained the requirements specified
by Procedure 40OP-9ZZ1 6.
/inclusions
Operator oversight and direction of the Unit 1 draindown to midloop evolution, and
decisions to take conservative actions during the evolution, were excellent.
Licensee
activities related to midloop operation demonstrated
a strong safety focus.,
01.2'e
r
ce Ta s
o e
Ins alled
Uni
1
pe
07
Condition report/disposition request (CRDR) 1-8-0142, dated March 16, 1998,
documented that a clearance acceptor walkdown of danger tagged equipment identified
components that were danger tagged, but not in the position required by the tags.
The
inspectors reviewed the circumstances surrounding this event to determine compliance
with applicable licensee procedures.
b.
se
a'ons and Fi di
s-
On March 16, 1998, tags for Clearances
98-01452, "Replace Solenoid Valve Type
V5H-71990," and 98-01074, "Replace Solenoid Valve Type V5H-71990," were hung and
independently verified by two licensed reactor operators.
Operators issued the
clearances to replace air-operated solenoid valves on Main Steam Isolation
Valves UV-0170 (SG 1, Line 1) and UV-0171 (SG 2, Line 1). As part of the clearance
boundary, electrical disconnects for the air solenoid valves were to be tagged in the OFF
position. The ON and OFF positions were clearly marked, with the ON position at the 12
o'lock position and the OFF position counterclockwise approximately 45 degrees.
At
. the time the operators hung the tags, other disconnect switches on the same panel were
tagged in the OFF position. Procedure 40DP-9OP30, "Clearance Processing," required
the clearance acceptor to perform a physical walkdown of the clearance boundary prior
to conducting work.
lt
-3-
During this walkdown, technicians (the clearance acceptor) observed that two electrical
disconnects on each clearance that were required to be tagged in the OFF position, were
tagged, but were in the ON position. As an immediate corrective action, the operators
involved were coached on the importance of adhering to the clearance procedure.
Investigation by the licensee determined that the root cause was a lack of attention to
detail on the part of the licensed operators.
Procedure 40DP-9OP30 provided the requirements for danger tag placement and
independent verification when a clearance was established.
Aweakness
in attention to
detail by two licensed reactor operators resulted in a violation of this procedure.
This
non-repetitive, licensee-identified, non-willfuland corrected violation is being issued as a
noncited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy
(50-528/9803-01).
ggnnclusi
ns
Inattention to detail by two licensed operators resulted in improper installation of danger
tags.
This was identified as an isolated example, and as a result, a noncited violation of
the clearance procedure was identified.
0 serv t'on of
ore Reloa
'
'
Ins
ion S o e
071
During this inspection period, the inspectors observed the licensee's performance of the
Unit 1 core reload as controlled by Procedure 72IC-9RX03, "Core Reloading,"
Revision 9.
Ob erva ion and Fi
in s
The Unit 1 core reload was appropriately controlled by Procedure 72IC-9RX03. Allfuel
movement was coordinated with the control room and logged in accordance with the
procedure.
Good communications were observed between the refueling senior reactor
operator, fuel pool bridge crane operator, and control room personnel.
On April4, 1998, the licensee experienced difficultyin seating slightly bowed Fuel
Assembly J511 in core Location B07. Procedural controls allowed the generation of a
transfer sequence
revision log to allow rotation of the fuel assembly
90'ounterclockwise
from the north to the west orientation.
The fuel assembly was
successfully seated in this orientation with the support from two other fuel assemblies.
Procedure 78OP-9FX01, "Appendix M - Action Plan for Movement of a Difficult
Assembly," Revision 9, recommended
that, ifdifficultywas encountered
in seating an
assembly,
a box of fuel assemblies was to be built around the difficultposition to provide
guidance for the difficultassembly.
On April 8, the fuel assembly was successfully
reoriented back to the north after the reload was complete.
~I
On April 8, operators suspended
core alterations and movement of irradiated fuel in
containment.
This was done as the result of identification of a cracked 3/8-inch stainless
steel sample line from Steam Generator 2, which indicated a potential containment
integrity issue (see Section M1.6 of this report). Operations allowed core alterations and
movement of irradiated fuel in containment to continue once the containment closure
requirements of Procedure 40ST-9ZZ08, "Containment Building Atmospheric
Penetrations Weekly Surveillance," Revision 1, were completed and verified to be
satisfactory.
~
c.
~Conclusi
The licensee demonstrated
good communications and a formal safety approach when
performing refueling operations.
Operations exercised good judgement by suspending
core alterations and movement of irradiated fuel in containment while a containment
integrity issue was being resolved.
02
Operational Status of Facilities and Equipment
02.1
Ins
cionof he
onro R omEs enial
ilra ion
ste
nis2and3
a.
I s ec ion Sco
e 71707
The inspectors performed an inspection of the control room essential filtration systems
for Units 2 and 3, using the following documents:
Updated Final Safety Analysis Report (UFSAR) Sections 6.4, 9.4, and 9.4.1
~
~ Preliminary Safety Analysis Report, Section 6.4
Technical Specifications, Section 3/4.7.7
Control Building HVAC [Heating Ventilation and AirConditioning] System Design
Manual, Revision 7
Procedure 33ST-9HJ01, "Control Room AirFiltration Unit AirflowCapacity and
Pressurization Test," Revision 2
Procedure 40OP-9HJ01, "Control Building HVAC Operations Procedure,"
Revision 4
Procedure 40ST-9HJ01, "Control Room Essential Filtration System Operability
Test," Revision 2
Procedure 16DP-OEP14, "Satellite Technical Support Center Actions," Revision 3
CRDR 2-6-0016
\\
l
-5-
CRDR 9-6-1415
CRDR 9-7-1961
bserva ions and Fin
i
s
During review of system documentation, the inspectors determined that the UFSAR
defined a limited stay time for personnel in the control room when the essential filtration
system was in the isolation mode.
The stay time was based on the number of people
and the expected carbon dioxide buildup level in the control room envelope.
The
inspectors identified that the licensee had not included this limitation in the appropriate
procedures to ensure control room habitability was maintained.
The licensee initiated a
CRDR to identify the appropriate corrective actions.
This item will be reviewed further to
determine what actions the licensee has taken to correct this issue and will be tracked as
an inspection followup item (50-529/9803-02).
The UFSAR stated that, while in the emergency mode, the essential filtration system was
designed to pressurize the control room to 1/4 inches of water gauge (iwg) pressure with
a makeup air flow of less than or equal to 1000 cfm. TS Section 3/4.7.7, the design
basis manual, and Procedure 33ST-9HJ01 stated that the required pressure was greater
than or equal to 1/8 iwg with makeup air flow less than or equal to 1000 cfm. The
inspectors questioned the licensee concerning the differences in the design and required
pressures
using the same flow rate.
The licensee stated that, in the original Preliminary Safety Analysis Report, dated
February 12, 1979, the required pressure was 1/4 iwg. However, NUREG 800,
Revision 2, issued in July 1981, reduced the required pressure to 1/8 iwg. Both
pressures
assumed
a maximum makeup air flow rate of 1,000 cfm. The makeup flow
rate was based on the ability of the control room charcoal filters to remove radionuclides.
The 1/4 iwg value in the UFSAR was the design capability, whereas, the 1/8 iwg value
referenced
in the TS and elsewhere was the design requirement.
The licensee was
performing an UFSAR review and planned on revising the appropriate sections to clarify
the difference between the two pressures.
The material condition and cleanliness of the systems was very good. The inspectors
observed that: (1) all accessible system components were in the required positions, (2)
power supplies and breakers were correctly aligned, (3) support systems, such as
essential chilled water and instrument air, were operational, and (4) system components
were correctly labeled.
c.
Conclusions
Maintenance of the control room essential filtration system was very good as evidenced
by the material condition of the components.
i
-6-
02.2
Wal down of h AFWS s
m Uni 2
a.
Ins ection Sco
e 71707
The inspectors conducted a walkdown of the Unit 2 AFW system to verify the as-built
configuration corlformed with the UFSAR description and piping and instrumentation
diagrams (P8 ID). The following P&IDs were reviewed:
P8 ID 02-M-AFP-001, "AuxiliaryFeedwater System," Revision 22
P&ID02-M-SGP-002, "Main Steam System," Revision 25
P8 ID 02-M-CTP-001, "Condensate Storage and Transfer System," Revision 16
b.
Observato sa dFindin s
The inspectors compared the UFSAR description of the AFW system to the Design Basis
Manual. The documents were in agreement on system design, function, and operation.
The inspectors walked down portions of the AFW system and verified the as-built
configuration conformed to the P&IDs. All major AFWvalves were in the required
position for full power operations.
A sample of key manual isolation valves in two
adjoining systems, the condensate
storage and transfer and main steam systems, were
verified locked open, as required by plant procedures.
In addition, the inspectors verified
that valve positions and indications for components with indications on both the control
room and remote shutdown panels were consistent with each other. The inspectors
identified no burnt out light bulbs or other material deficiencies on either of the control
panels.
The material condition of the AFW system was good.
c.
conclusions
The Unit 2 AFW system was installed and maintained per the applicable design and
operating requirements for those portions of the system reviewed. The material
condition of the AFW system was good.
02.3
alkdownoftheConta'nmentS
ra
S sem
Unit3
a.
Ins ec ion Sco
e 71707
The inspectors reviewed the Unit 3 CS system to verify the UFSAR description
conformed to the as-built condition, Design Basis Manual, and the individual plant
examination (IPE) assumptions.
The inspectors also verified the licensee had
adequately addressed'motor-operated
valve pressure locking and thermal binding
concerns. The inspectors also reviewed the unavailability of the system.
The following
documents were reviewed:
~
Design Modification Work Order (DMWO) 00768975, "Design to Eliminate
Pressure
Locking Thermal Binding Problems in Sl Valves (Sl-671 8 672)"
-7-
Mechanical System Drawings M-SIP-001, Revision 22; M-SIP-002, Revision 19;
and M-SIP-003, Revision 07, "Piping and Instrumentation Diagram Safety
Injection 8 Shutdown Cooling System"
The inspectors also walked down the CS system and interviewed the system engineer
and control room operators.
bserv
i n
d Findin s
The inspectors compared the UFSAR description of the CS system to the Design Basis
Manual and IPE assumptions.
In general, the documents were in agreement on system
design, function, and operation.
Minordiscrepancies
noted by the inspectors had also
been identified by the licensee as part of the UFSAR review program.
The identified
discrepancies
did not effect system operability. The IPE Human Reliability Analysis
section detailed the human performance issues associated with the CS system.
The
inspectors reviewed emergency operating and annunciator response procedures and
verified these documents adequately addressed
the IPE human response
issues.
The inspectors reviewed DMWO 00768975, which was generated to address the
concerns of NRC Generic Letter 95-07, "Pressure Locking and Thermal Binding of
Safety-Related Power Operated Gate Valves." The licensee performed an analysis and
concluded that CS Header Isolation Valves SIBUV-0671 and SIAUV-0672 were
potentially susceptible to Generic Letter 95-07 concerns.
Corrective actions were
initiated by the licensee to eliminate this susceptibility. The DMWO, which increased the
capacity of the valve motor operator to overcome the potential thermal binding of the
valves, included the required 10 CFR 50.59 screening and evaluation.
The evaluation
addressed
the effects on the system design because of the added power consumption
on the EDG and a slight change in valve stroke time and concluded these changes to the
actuator were within the limits of the UFSAR. The design modification package was well
prepared and complete, including the 10 CFR 50.59 evaluation.
The licensee had completed the modification on Valve SIBUV-0671 in Units 2 and 3, and
was performing the modification in Unit 1 during the refueling outage.
The modifications
of Valve SIAUV-0672 were to be performed in each unit during their next refueling
outage.
This time line was consistent with the licensee's implementation schedule for
corrective actions, which was submitted to the NRC by letter dated June 28, 1996.
The inspectors discussed, with the system engineer, the unavailability of the CS system
and reviewed performance and condition monito'ring activities associated with the
system.
The system engineer provided documentation indicating that the unavailability
of the CS systems for all three units averaged approximately
1 percent over the last year
and was well below the maintenance
rule trigger level of 3.75 percent unavailability.
Further, the low number of maintenance
rule functional failures compared to the number
of demands. on the system (1/450), over the last 4 years, demonstrated the CS system
far exceeded the performance criteria mandated by the Maintenance Rule Program.
The
system engineer was very knowledgeable of the system.
1
fl
'1
1
ll
h
-8-
The inspectors conducted a walkdown of portions of Unit 3 CS Train A and verified that
the as-built configuration conformed to P8 IDs. All observed valves were in the required
position for full power operations.
The material condition of the CS system was good.
Some minor material condition and housekeeping
issues were identified and brought to
the'attention of the control room shift supervisor.
These concerns were addressed
and
corrected in a timely manner.
None of the concerns affected operability of the CS
system.
The inspectors also verified operability of the CS pumps, motor-operated valves, check
valves, and relief valves in the system design basis accident flow path.
Operability was
determined by ensuring that these components were included in the surveillance testing
program and that the tests were current. The inspectors further verified that these
surveillances were performed within the required periodicity for the last 18 months.
Conclusions
Unit 3 CS Train A was installed and maintained per the applicable design and operating
requirements for those portions of the system reviewed.
The material condition of the CS
system was good. The CS system unavailability was very low.
08
Miscellaneous 0 erations Issues
92901
08.1
Closed
Viola 'on 50-529/9716-01:
Inadequate Shutdown Cooling Ftow
This violation documented that Unit 2 was operated in Mode 5 with less than the required
shutdown cooling flow. The inspectors reviewed the response,
dated
November 26, 1997, to the Notice of Violation and verified that the corrective actions
were appropriately implemented to prevent recurrence.
II. Maintenance
M1
Conduct of Maintenance
M,1.1
General Commen so
Maintenance Ac iviies Units 1
2 and 3
a.
Ins ection Sco
e 62707
The inspectors observed all or portions of the following activities performed per the
following work orders (WO):
WO 828772:
Rework the Seat Leakage and Repack Valve 1PSIBV842 As Required
WO 828802:
Disassemble Valve 1PSIBV555 and Rework As Required To Stop
Leakage Past the Seat.
Replace the Downstream Nipple and Cap.
-9-
WO 803415:
ModifyValve 1JSIBUV0667 From a Rising, Rotating Stem To a Rising,
Non-rotating Stem', and Replace Operator.
WO 755377:
Implement The Electrical Portion of DMWO 739596 (Design Package For
Low Pressure Safety Injection Pump Shaft Retrofit) Per The Attached
Instructions (Unit 1).
WO 835433:
Troubleshoot Diesel Control Circuits Per Engineering Gameplan (Unit 2).
WO 835620:
Lubricate and Functionally EDG Fuel Cylinder Control Valves (Unit 2).
WO 082703:
EDG B Piston Modification.
WO 081649:
WO 716930:
Replace Capacitors in Inverter 1-E-PKC-N43.
Install Transformer Cooling Fan Units in Accordance With DMWO 704966
On 4160-vac to 480-vac Transformer 1E-PGA-L31.
WO 767832:
Install Bonnet Bypass Spring Check Valve 1JAFCUV0036.
WO 803414:
Install DMWO 705815 To Install SMB-00-10 Limitorque Operator On
1JSIBUV0646, High Pressure Injection System 2 Flow Control To RC 1B
Containment Isolation.
b.
Observa 's
n
Findi
The inspectors found the work performed under these activities to be professional and
thorough.
Allwork observed was performed with the work package present and in active
use.
Generally, good work practices were observed.
practices were good. Technicians were experienced and knowledgeable of their
assigned tasks.
c.
Conclusions
Routine and outage-related
maintenance activities were generally conducted in a safety
conscious manner by knowledgeable technicians using approved procedures.
Good
work and foreign material control practices were observed.
M1.2
General
ommen
on S
'Ilanc
Activities Uni 3
a.
Ins ec ion Sco
e
1726
The inspectors observed portions of Surveillance Test 73ST-9CL10, "Containment
Ventilation Purge Isolation Valves (42 inch) - Penetration 57 (Unit 3)."
4
i
i
1'
-10-
Observa ions and Findin s
The inspectors found this surveillance to be performed acceptably and as specified by
the applicable procedure.
M1.3
Failure to Perform All Prere
uisites Prior o Ins ectin
In ernals of an ED
ni 1)
Ins ec ion Sco
e
62707
During the Unit 1 refueling outage, the inspectors observed maintenance
being
performed on EDG A. The inspectors reviewed associated
procedures, work orders, and
active clearances.
b.
Observa ion
n
Fin in
The licensee was performing Procedure 31ST-9DG01, "Diesel Engine Surveillance
Inspection," Revision 16, during the refueling outage.
The objective of the procedure
was to perform the annual and 5-year inspection requirements,
as specified in Vendor
Technical Manual VTM-C628-001. The purpose of the procedure was to perform and
document inspections, analysis, and evaluations to ensure the mechanical condition of
the EDG was maintained in agreement with the manufacturer predictive maintenance
guidelines.
On March 31, 1998, the inspectors observed mechanics perform boroscope inspections
of the cylinder liners through the injector ports in accordance with Step 8.7 of
Procedure 31ST-9DG01.
The inspector identified that Prerequisite Steps 7.4.1 and 7.11,
- which provided authorization for work activities to begin, had not been completed prior to
the start of work. The licensee halted boroscope inspections and removed the
boroscope.
Appropriate signatures documenting the completion of all prerequisites were
obtained prior to recommencement
of work. The licensee initiated CRDR 1-8-0206 to
document this event.
Personnel failed to obtain the appropriate authorization prior to commencing boroscope
inspections, as required by procedure.
The failure to follow a procedure for a
safety-related work activity is a violation of TS 6.8.1 (50-528/9803-03).
Recent NRC Inspection Reports 50-528;529;530/98-02
and 50-528;529;530/97-018
document varying degrees of work performance problems in the maintenance
area.
One
report documented
a weakness where a work order was inadvertently misplaced while
work was in progress.
Another report documented that, during the performance of a
surveillance task, documentation was not being kept current.
Management recognized
the significance of the recent failures and was reiterating their expectations on proper
procedural adherence
to the maintenance staff.
ji
-11-
~
Engineering personnel performed boroscope inspections on Unit 1 EDG A cylinder liners
without obtaining proper authorization.
This is a violation of TS 6.8.1 for the failure to
follow procedure.
M1.4
ov m n S en Fue
ol
i
ns
i
co
617 6
CRDR 9-8-0511, initiated on March 27, 1998, documented an event that occurred when
refueling and mechanical services personnel relocated a spent fuel bundle inside the
Unit 1 spent fuel pool. During movement of the bundle, higher than expected radiation
levels were experienced on the spent fuel handling machine and locally near the spent
fuel pool on the 140-foot elevation.
The inspectors reviewed the circumstances of this
event to determine ifcompliance with applicable licensee procedures was a contributing
cause.
b.
a
Findin s
On March 26, refueling and mechanical services personnel planned and conducted three
tasks inside the Unit 1 fuel building. One of these activities was to move a fuel bundle
from one location to another inside the spent fuel pool.
The target location for the bundle was directly in front of and approximately 2 feet from
the steel gate separating the fuel transfer canal and the spent fuel pool. The spent fuel
pool was at its normal level of 137 feet.
However, the fuel transfer canal was being kept
at a lower level to minimize the amount of leakage past the Fuel Transfer Tube Isolation
Valve PCN-V118. Prior to starting the job, the team conducted a briefing in the control
room with the shift manager.
Radiation protection personnel were also notified of the
jobs and their scope.
With the level in the transfer canal low, the extra shielding that
would have been provided by the water in the transfer canal did not exist.
Refueling and
mechanical services personnel discussed this with the shift manager and it was decided
that an adequate
level existed in the fuel transfer canal to support the fuel move.
This judgment was made based upon the guidance provided in Procedure 78OP-9FX03,
"Spent Fuel Handling Machine," Revision 8. When the technicians moved the bundle, it
was moved in such a manner that it approached the face of the gate in a vertical
orientation. As the bundle neared the gate, radiation levels on the spent fuel handling
machine and outside of the fuel pool began to rise. The levels exceeded the 15 mr/hr
alarm set point of the area monitor on the spent fuel handling machine and caused a
frisker near the fuel pool to alarm. An RP technician in the area measured
a field of
approximately 300 mr/hr near the fuel transfer canal. The trainee who was operating the
spent fuel handling machine under instruction was directed to leave the area and the
experienced operator continued operation of the spent fuel handling machine.
At the
time of the radiation alarms, the bundle was above the target location and, by direction of
i
0
-12-.
an RP technician, was lowered into the fuel racks to clear the condition. The elevated
radiation levels existed for less than
1 minute.
Due to the short duration, no significant
personnel exposure resulted.
The licensee initiated CRDR 9-8-0511 to document this
event.
The immediate corrective action was to require that the fuel transfer canal level
be within 6 inches of spent fuel pool level prior to any fuel movement.
Operation of the spent fuel handling machine was conducted in accordance with
Procedure 78OP-9FX03, which listed the prerequisites
in Section 4.2. Step 4.2.1
directed personnel to check the level in the cask loading pit and fuel transfer canal.
Although the level in the fuel transfer canal was considered to be satisfactory by the shift
manager prior to making the fuel move, adequate compliance with this prerequisite step
was not achieved because of the inadequate guidance provided in the procedure.
Specifically, the procedure did not provide sufficient information on how to determine
what the minimum level should be. The failure to adequately maintain a procedure is a
violation of TS 6.8.1 (50-528/9803-04).
~Conc
sions
M1.5
A violation was identified for the failure to provide specific criteria for licensee personnel
to determine the minimum required level in the transfer canal prior to moving spent fuel.
This is a violation of TS 6.8.1 for the failure to provide an adequate procedure.
a
in S
am
enera
2 Dow co
er Blowdown Sam
le Line
ni
1
nsecin
c
e 62
7
The inspectors evaluated the licensee's actions to repair a leak on a 3/8-inch diameter
downcomer sampling line connected to Unit 1 Steam Generator 2. The applicable
Deficiency Work Order 00775579 and supporting calculations were reviewed.
Observa io s and Findin s
In October 1996, the licensee performed a temporary Furmanite leak seal repair to stop
a leak identified on Sample Line 1PSGEL181A for Steam Generator 2. The licensee
intended to permanently repair the line to return it to its original configuration during the
current Unit 1 refueling outage.
While removing the Furmanite box from the sampling
line, the licensee observed a crack in the Swagelok fitting adjacent to the steam
generator nozzle connection for the sampling line. There was still a short section of
tubing and a ferrule protruding out of the existing threaded male fitting welded to the
To keep the design function of this tubing configuration intact, maintenance engineering
repaired the line by the following method. A socket-welded tubing union was welded to
the non-threaded
side of a Swagelok nut. The nut,was threaded onto the existing fitting
welded to the steam generator.
Once in place, the threaded assembly was seal welded
to ensure a leak tight joint. The stainless steel tubing was socket welded to the
I
-13-
socket-welded side of the Swagelok fitting assembly.
The installation was consistent
with the licensee's design requirements,
including the ASME code.
The inspector
considered that by seal welding the threaded joint, the licensee had taken additional
conservative measures to ensure system leak tightness.
The inspectors reviewed the evaluation completed by the licensee prior to installing the
temporary Furmanite leak seal.
Engineering determined that additional restraint of the
system would be required to accommodate
the additional weight of the Furmanite box.
The decision was made to attach the Furmanite box to an adjacent box connected to the
This was accomplished by using long bolts that protruded from the
Furmanite box and were tack welded to the adjacent box. The WO specified that the
tack weld be made at each of the four protruding bolts and that the welder should
attempt to perform the tack weld around as much of the circumference of each protruding
bolt as possible.
The inspectors noted that tack welds are typically used by welders to
hold pieces into place for welding, and tack welds do not have any particular
requirements that could be converted into an equivalent structural strength.
Because of
the small loads associated with this modification, however, the instructions to tack weld
would be within the skill of the craft to accomplish a structurally sufficient weld. This was
supported by the licensee's observation that the welds were intact when the Furmanite
box was removed and the line was returned to its original configuration.
In reviewing the stress model of the sampling line, the inspectors noted that a 3-way
restraint had been modeled at the location of the Furmanite box. The inspectors
disagreed with this modeling of the restraint because the Furmanite box was securely
tightened onto the 3/8-inch sampling line and would not have allowed it to rotate.
The
inspectors considered that modeling an anchor that would restrain the line from moving
in all three translational and rotational directions would have been appropriate.
The
licensee remodeled the system considering an anchor-type restraint at the Furmanite
box. Although some stresses
and loads increased
in the ren>odeled system, the system
did not exceed its design basis.
Conclusion
The permanent repair of a leak in the Unit 1 Steam Generator 2 downcomer sampling
line was consistent with system design requirements.
M2
Maintena'nce and Material Condition of Facilities and Equipment
M2.1
Review of Material Condi ion Durin
Plan T
ns
e
ion Sco
e 62707
e
During this inspection period, routine tours of all units were conducted to observe the
status of plant equipment and evaluate plant material condition.
<J
-14-
rva ion
indin s
Inspector observation of plant material condition identified no major observable material
condition deficiencies.
Minor deficiencies brought to the attention of the licensee were
documented with work requests.
During the Unit 1 outage, the inspectors observed several disassembled
electrical
components and motor-operated valve actuators.
This included 125-vdc to 120-vac
Inverter PNC-N13, 125-vdc to 480-vac Inverter PKC-N43, 4160- to 480-vac Transformer
PGA-L31, Containment Electrical Penetration 1ESFBZ380, and high pressure safety
injection flow control Valve Sl-UV-646. The material condition of these components was
good.
c.
~Con ~vien
Observable material condition of the three units was satisfactory.
Material condition of
the interior of components disassembled
for the Unit 1 outage was good.
M2.2
Con
in
en
lo
re
al
wn
ni
1
a.
I
ion coe 670
Tours of the Unit 1 containment were made to assess
material condition of the
equipment.
b.
0 s rvaionsa
dFindi
s
On April 15, 1998, with the unit in Mode 5, the inspectors accompanied the licensee
during performance of Procedure 40ST-9ZZ09, "Containment Cleanliness Inspection,"
Revision 9. Prior to the tour, the inspectors reviewed CRDR 98-0620, which
documented
a list of materials inside the containment building. These materials were
required for inservice inspection at normal operating pressure and temperature
(NOP/NOT) during Modes 4 and 5. Containment cleanliness and material condition were
generally good. The inspectors identified minor debris in various areas, which the
licensee retrieved and disposed of properly. The containment coordinator indicated that
an additional tour would be done at NOP/NOT, at which time they would verify final
containment cleanliness.
On April 17, the inspectors conducted an independent walkdown of portions of the
containment at NOP/NOT. Canned lagging that had not been installed at the time of the
previous inspection was identified and properly staged for reinstallation.
The inspectors
did not observe any active leakage or evidence of previous leakage.
I
-15-
c.
~Con Iusions
Material condition of the equipment in the Unit 1 containment was good.
MS
Miscellaneous Ilaintenance Issues
M8.1
I
Lic nsee Even
e o
E
5 -528/96-003-01: Open AuxiliaryBuilding Door
Causes
Fuel Building Essential Filtration Inoperability.
Revision 0 of this LER noted problems with door numbering ergonomics and door control
methods that contributed to the event.
Details of the event were discussed
in NRC
Inspection Report 50-528, -529, -530/96-13.
Revision
1 of this LER described the
following additional corrective actions implemented by the licensee to prevent
recurrence:
New door signage clearly differentiated between the door number and room
number.
Procedure 40DP-9ZZ17, "Control of Door Hatches, and Floor Plugs;" was revised
(Revision 8) to update the design basis functions of the doors.
~
Training on procedure changes and door control was accomplished.
The inspector verified these corrective actions to be acceptable and complete.
E1
Conduct of Engineering
E1.1
Review of Modifica io
o
n Ins ection Port o Steam Generator
1
Unit 1)
cin
co e 3751
The inspectors reviewed the licensee's
10 CFR 50.59 screening and evaluation for
adding a 2.25-inch inspection port to Unit 1 Steam Generator 1.
b.
Observations and
Find'uring
routine eddy-current testing inspection of Steam Generator 1, degradation of
several tubes just above the cold leg flow distribution plate was observed.
It appeared
that the degradation was potentially caused by foreign objects in the vicinity. The
proposed inspection port would allow the capability to visually inspect and potentially
retrieve any foreign objects from this area.
The port was designed to extend through the
5.75-inch thick secondary side shell and through the nominal 1.5-inch thick economizer
feedwater box.
It was located approximately 4 inches above the elevation of the top of
the flow distribution plate.
Four 3/4-inch diameter stud holes and a gasket mating
4
I
-16-
surface were machined into the secondary side shell to allow closure of the inspection
port by a gasketed cover plate.
The inspectors noted that the 10 CFR 50.59 screening and evaluation were thorough.
Justifications to the generic screening questions were comprehensive
and detailed.
Changes to the UFSAR were identified and it was determined that the modification was
not an unreviewed safety question.
c.
~Conclusi
The licensee's
10 CFR 50.59 screening and evaluation for the modification to add an
inspection port to Unit 1 Steam Generator
1 was thorough and comprehensive..
E1.2
Year 2000
Y2
eadiness
Un'ts1
2 and 3
a.
ns
c ion
c
e
37551
On March 26, 1998, the inspectors conducted discussions concerning the licensee's
readiness to address operational and equipment problems resulting from the inability of
some components and computer software to correctly interpret some dates that occur
before and after the year 2000.
b.
Observa ions and Findin s
Computers and related technologies are in use throughout the equipment in the three
units and by plant support organizations.
As a result, issues raised by the Y2K problem
require action on the part of the licensee to ensure continued safe operation of the units.
The potential problems related to the Y2K issue include possible simultaneous trip of all
three units and possible unavailability of emergency power sources.
To address the Y2K problem, the licensee established
a project organization, in
January 1998, which reported directly to the Vice President, Nuclear Engineering.
It had
a full-time project manager and a core team of individuals from different disciplines. Also
associated
with the project were single points of contact in key areas.
The project was
proceeding in a planned manner to assess
the extent of the Y2K problem, determine the
extent of remediation necessary,
and correct the identified problems.
There were four areas that were being addressed
by the Y2K project, as discussed
below:
The first involved problems with programmed computer chips internal to some
components; referred to as embedded systems.
Examples of components with
possible embedded systems were certain transmitters and data recorders.
The
licensee was approximately 80 percent complete in their efforts to inventory
embedded systems.
The licensee expected to have completed the assessment
of the embedded system components by October 1998.
>i
f>
b
-17-
~
The second area involved process systems, such as the core protection
calculators.
The licensee expected to have a detailed, Y2K remediation plan
completed by June 1998.
~
The third area involved information technology supported systems, such as large
database
systems and engineering computer code.
The licensee expected to
complete items in this area by December 1998.
The fourth area involved department supported systems, such as the plant
simulation facilityand computerized measuring and test equipment.
The licensee
expected to have a detailed Y2K remediation plan completed by June 1998.
c.
Conclusions
Licensee plans to address operational and equipment problems resulting from the
inability of some components and computer software to correctly interpret some dates
that occur before and after the year 2000 appeared
to be comprehensive.
IV
a
Su
R1
Radiological Protection and Chemistry Controls
R11
Gen
r
I
mmenso
nr Is
is1
d3
a.
In
ection Sco
e
17 0
The inspectors monitored RP activities during routine site tours.
b.'bserva
ions and Findin
The inspectors observed radiation protection personnel, including supervisors, routinely
to'uring the radiologically controlled areas.
Licensee personnel working in radiologically
controlled areas exhibited good radiation work practices.
Contaminated areas and high radiation areas were properly posted.
Area surveys
posted outside rooms were current. The inspectors checked a sample of doors, required
to be locked for the purpose of radiation protection, and all were in accordance with
requirements.
Conclusions
The radiological protection program was effectively implemented in those ares reviewed.
I
,l
-18-
V
a
a
e
entMeeti
s
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of the licensee's staff at the
conclusion, of the inspection on April 21, 1998. The licensee acknowledged the findings
presented.
a
The inspectors asked the licensee whether any material examined during this inspection
should be considered proprietary.
No proprietary information was identified.
c
PARTIALLIST OF PERSONS CONTACTED,
P. Borchert, Shift Manager, Operations
D. Garnes, Unit 2 Department Leader, Operations
P. Crawley, Director, Nuclear Fuel Management
D. Fan, Section Leader, System Engineering
D. Kanitz, Engineer, Nuclear Regulatory Affairs
P. Kirker, Unit 3 Department Leader, Operations
A. Krainik, Department Leader, Nuclear Regulatory Affairs
J. Levine, Senior Vice President, Nuclear
D. Mauldin, Director, Maintenance
D. Marks, Section Leader, Nuclear Regulatory Affairs
M. Muhs, Department Leader, RAMS/Maintenance
G. Overbeck, Vice President, Nuclear Production
T. Radke, Director, Outages
J. Scott, Director, Site Chemistry
G. Shanker, Department Leader, Speciality Engineering
M. Shea, Director, Radiation Protection
D. Smith, Director, Operations
E. Sterling, Department Leader, Nuclear Assurance Department
J. Velotta, Director, Training
P. Wiley, Department Leader, Operations
M. Winsor, Department Leader, Maintenance Engineering
INSPECTION PROCEDURES USED
IP 37551:
IP 60710:
IP 61726:
IP 62707:
IP 71707:
IP 92901:
Onsite Engineering
Refueling Activities
Surveillance Observations
Maintenance Observations
Plant Operations
Plant Operations Followup
i
I
0
~Oned
50-'528/9803-01
50-528/9803-02
-2-
ITEMS OPENED CLOSED, AND DISCUSSED
Improper installation of da
o danger tags (Section 01.2)
IFI
Review licensee's actions for con
ons or control room habitability limi s
50-528/9803-03
EDDG internal inspections performe
(S
t
M1 3)
Failure to provide an adequate procedur
(S t'1 4)
50-529/9716-01
50-528/96-003-01
50-528/9803-01
Inadequate shutdown c
n cooling flow (Section 08.1)
LER
Open auxilia
'
ry building door causes fuel b '
bilit (S
tio
M8.1)
Improper installation of da
o danger tags (Section 01.2)
S 0
US
D
ARN
CFR
CRDR
DMWO
. IPE
Iwg
LER
NOP/NOT
P8 ID
TS
Auxiiliaryfeedwater
Code of Federal Regulations
p rt/disposition request
.
Condition re o
ontainment spray
Design modification work o d
gency diesel generator
or er
Heating ventilation and air conditionin
inches ofwater gauge
Licensee event report
normal operating pressure and tern era
Piping and instrume
t
t'enation diagram
Radiation protection
Technical Specifications
Safety Analysis Report
Updated Final S
'
I