ML14181A729
| ML14181A729 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/17/1995 |
| From: | William Orders, Verrelli D NRC Office of Inspection & Enforcement (IE Region II) |
| To: | Carolina Power & Light Co |
| Shared Package | |
| ML14181A726 | List: |
| References | |
| 50-261-95-19, NUDOCS 9507240360 | |
| Download: ML14181A729 (20) | |
See also: IR 05000261/1995019
Text
pf REGUo
UNITED STATES
o
NUCLEAR REGULATORY COMMISSION
REGION II
4 7
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report No.:
50-261/95-19
Licensee:
Carolina Power and Light Company
P. 0. Box 1551
Raleigh, NC 27602
Docket No.:
50-261
License No.:
Facility Name: H. B. Robinson Unit 2
Inspection Conducted:
May 14 - June 17, 1995
Lead Inspector:
7-f 7___ 5_
T
T. Orders, Senior Resident Inspector
Date Signed
Other Inspectors: C. R. Ogle, Resident Inspector
P. J. Fillion, Reactor Inspector
W. Garne , Project Engineer
Approved by:
_
d__
___'____--_
__7-/7-9_
Dav'd M. Vefelli, Chief
Date Signed
Rea tor Projects Branch 1A
Division of Reactor Projects
SUMMARY
SCOPE:
This routine, resident inspection was conducted in the areas of plant
operations, maintenance activities, engineering efforts, and plant support
functions. The inspection effort included reviews of activities during non
regular work hours on May 14, 17, 21, 23, 30 and June 1, 2, 3, 4, 8, 14, and
17, 1995.
RESULTS:
Plant Operations [Paragraph 3]:
A violation was identified involving multiple examples of configuration
control events.
An unresolved item was identified involving racking-in of an SI pump breaker
with LTOPP in service.
A second unresolved item was identified involving loose paint in containment.
An improving trend in the material condition of components and structures in
the Auxiliary Building was noted.
9507240360 950717
PDR ADOCK 05000261
Q
2
Maintenance [Paragraph 4]:
A violation was identified involving an inadvertent RHR pump start during
maintenance.
A non-cited violation was identified involving personnel not following a FMEA
procedure.
Engineering [Paragraph 5]:
A non-cited violation was identified involving the licensee's failure to
incorporate load sequencing timer settings into appropriate design documents.
In the main, the licensee's performance in implementing the control room human
factors enhancement modification was good. However, the safety evaluation for
the modification did not accurately describe the effects of deletion of two
non-safety-related annunciator points from the control room. This fact
represents a weakness in the design control process.
The failure to identify a potentially intermittent abnormal auxiliary
feedwater pump sequencing response during surveillance testing was considered
a weakness.
REPORT DETAILS
1.
PERSONS CONTACTED
Licensee Employees:
W. Brand, Supervisor, Environmental and Radiation Control
- M. Brown, Manager, Design Engineering
- P. Cafarella, Superintendent, Mechanical Systems
- A. Carley, Manager, Site Communications
- B. Clark, Manager, Maintenance
D. Crook, Licensing/Regulatory Compliance
- A. Garrou, Acting Manager, Licensing/Regulatory Programs
D. Gudger, Senior Specialist, Licensing/Regulatory Programs
- M. Herrel, Manager, Training
- C. Hinnant, Vice President, Robinson Nuclear Project
P. Jenny, Manager, Emergency Preparedness
- K. Jensen, Supervisor, Reactor Systems
- M. Knacszck, Superintendent, Projects
J. Kozyra, Licensing/Regulatory Programs
- R. Krich, Manager, Regulatory Affairs
E. Martin, Manager, Document Services
- B. Meyer, Manager, Operations
- G. Miller, Manager, Robinson Engineering Support Section
- J* Moyer, Manager, Nuclear Assessment Section
- P. Musser, Manager, Plant Operations Assessment
W. Randlett, Manager, Security
B. Steele, Manager, Shift Operations
- R. Stewart, Robinson Engineering Support Section
- W. Stover, Manager, Operations Procedures
D. Taylor, Plant Controller
G. Walters, Manager, Support Training
R. Wardern, Manager, Plant Support Nuclear Assessment Section
W. Whelan, Industrial Health and Safety Representative
- D. Whitehead, Manager, Plant Support Services
T. Wilkerson, Manager, Environmental Control
L. Woods, Manager, Technical Support
- D. Young, Plant General Manager
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
NRC Personnel:
- W. Orders, Senior Resident Inspector
- C. Ogle, Resident Inspector
- P. Fillion, Reactor Inspector
- L. Garner, Project Engineer
- Attended one or more of the three exit interviews conducted for this
report necessitated by visiting RH inspectors.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. PLANT STATUS AND ACTIVITIES
a.
Operating Status
The report period began with the unit in day 16 of refueling
outage 16.
Following the completion of planned outage work, with
the unit at normal operating temperature and pressure, an RCS leak
was identified on the main flange area of the C reactor coolant
pump. This forced a plant cooldown to conduct repairs. Following
these repairs, an orderly transition was made through plant fill,
heatup, and startup. The unit output breakers were shut on
June 21, 1995, day 54 of the outage.
b.
Other NRC Inspections and Meetings
P. Fillion, a Region II Reactor Inspector, was on site during the
week of May 22 - 26, 1995, to conduct an inspection of
modifications to the control room. Results of this inspection are
contained in this report.
L. Garner, a Region II Project Engineer, was on site during the
week of June 12 - 16, 1995, to conduct an inspection of station
modifications and major surveillance testing. Results of this
inspection are contained in this report.
3. OPERATIONS
a.
Plant Operations (71707)
The inspectors evaluated licensee activities to determine if the
facility was being operated safely and in conformance with
regulatory requirements. These activities were assessed through
direct observation, facility tours, discussions with licensee
personnel, as well as, management, evaluation of equipment system
status, and review of facility records.
Routine plant tours were conducted to evaluate equipment
operability, assess the general condition of plant equipment, and
to verify that radiological controls, fire protection controls,
physical protection controls, and equipment tagging procedures
were properly implemented. During routine inspections of the
Auxiliary Building by a Region II inspector, it was noted that the
external material condition of plant equipment had improved as
compared to that observed approximately two years ago. This
observation was based upon fewer components, such as valves, with
boron acid buildup due to leaks; leak catch container.usage has
become infrequent; and equipment and structural coatings (paint)
have been improved.
3
Clearance Procedure Error, Valve SI-883R
On May 26, 1995, the licensee experienced difficulties filling the
safety injection accumulators. The licensee determined that valve
SI-883R was shut with a clearance tag attached. This valve
isolates the accumulator fill header from the SI pump discharge
flowpath and is normally open. The tag was removed, the valve was
restored to the proper position, and a condition report was
generated. The inspectors were informed that the clearance tag
hanging on the valve was from a local clearance and test request
which had been canceled on May 2, 1995.
The inspectors interviewed the Operations personnel involved in
the disposition of the clearance tag found hanging on SI-883R,
reviewed all clearances identified as having been on the valve
during the current refueling outage, and reviewed licensee
procedures: Operations Management Manual Procedures, OMM-005,
Clearance and Test Request; OMM-001, Operations - Conduct of
Operations; and Plant Program Procedure, PLP-30, Independent
Verification.
The inspectors determined that an auxiliary operator and an
independent verifier initialled the clearance on May 2, 1995,
indicating that the valve was open and the tag removed.
Subsequently, a licensed senior reactor operator signed the
clearance attesting that all tags listed in the clearance were
accounted for. In fact these activities had not been accomplished
for valve SI-883R.
The AO and independent verifier stated that on May 2, they entered
the CV to both remove and install several clearances and that
working copies of the clearances were taken into the CV to
accomplish these activities. The AO stated that SI-883R was not
repositioned when the clearance tag was removed since other
clearance tags were attached to the valve which required that it
remain shut and he thought that he denoted that fact on the
working copy of the clearance. These individuals also informed
the inspectors that upon exiting the CV, the working copies of the
clearances and the tags were discarded as potentially contaminated
material. Contrary to the requirements of OMM-005, the tags which
had been removed were not "called-in" to the clearance center for
accountability before they were disposed of. When the AO and
independent verifier returned to the work control center, they
completed the master copy of the clearance from memory indicating,
in part, that SI-883R had been opened and the tag removed.
The inspectors interviewed the SRO who signed the clearance
attesting to the fact that all tags and caps associated with it
had been accounted for. He stated that he did not recall the
specifics of the clearance in question, but speculated that the
clearance may have been removed incrementally. He stated that
when clearances are removed in this fashion, caps and tags are not
4
retained until the clearance is completely removed and hence, no
final verification of tag accountability is performed. He stated
that in this situation, the licensed operator in the clearance
center relies on the initials on the clearance as the basis for
verification of accountability.
The inspectors concluded that the individuals involved in the
restoration of SI-883R failed to comply with the requirements of
OMM-005. This is identified as one of six examples which
collectively constitute a violation, 50-261/95-19-01: Operations
Configuration Control Events Concerning RHR Pump Flow Path,
SI-883R, Steam Driven Auxiliary Feedwater, And Containment
Ventilation Unit.
HVH-2 Run With Air Flowpaths Isolated
On May 26, 1995, during a tour of containment, the inspectors
noted an abnormal noise coming from containment recirculation fan
unit, HVH-2. The unit was running but the inlet damper and intake
butterfly valve were both closed. The inspectors notified the
control room and the unit was stopped. A condition report was
generated to address this issue.
The inspectors interviewed the control room SRO and two SROs
assigned to the clearance center at the time, reviewed Local
Clearance And Test Request 95-FO476 which was in force on the HVH
unit at the time of this observation, and evaluated Operations
Management Manual Procedure, OMM-005, Clearance And Test Request.
The inspectors determined that the inlet damper to HVH-2 was
failed closed as a result of instrument air supply to the damper
activator being isolated by clearance 95-FO476. The clearance did
not alter the position of the butterfly valve or its air supply
valve. The clearance specified that a CIT be affixed to the RTGB
control switch for HVH-2 to alert operators of the clearance. The
inspectors were informed that a CIT was not on the switch when the
unit was started. From a review of the clearance paperwork, the
inspectors noted that no signature or initials were recorded to
demonstrate that the CIT had been affixed.
The inspectors concluded that the failure to affix the CIT was
contrary to the requirements of OMM-005. This example constitutes
one of six examples which collectively comprise Violation 50-261/
95-19-01, Operations Configuration Control Events Concerning RHR
Pump Flow Path, SI-883R, Steam Driven Auxiliary Feedwater, And
Containment Ventilation Unit.
5
SI Pump Breaker Racked In With LTOPP In Service
At 5:26 a.m., on the morning of May 30, 1995, the A SI pump motor
breaker was racked in to fill the SI accumulators. Approximately
one minute later, the RCS vent path to containment was isolated
when the pressurizer PORVs were unblocked and shut in preparation
for placing LTOPP in service. At 5:31 a.m., this activity was
complete and LTOPP was declared in service. This configuration
existed until 6:08 a.m., when the SI pump breaker was again
racked-out.
Having an SI pump breaker racked in with the RCS not vented
appears to be contrary to licensee procedures and TS 3.3.1.3.
Pending a review of the licensing basis associated with TS 3.3.1.3
and LTOPP, this will be tracked as an Unresolved Item,
URI 50-261/95-19-02, SI Pump Breaker Racked-In With LTOPP In
Service.
RHR Pump Operated With No Flow
On June 3, 1995, the licensee was preparing to restart the unit,
having completed refueling. Using GP-002, Cold Shutdown To Hot
Subcritical At No Load TAVG, control room operators were
performing procedure steps to depressurize and cooldown the "A"
train of the RHR system after having isolated it from the Reactor
Coolant System. This is done by recirculating the RHR train
through its associated heat exchanger until it has been cooled
down to approximately 150* F. After approximately fifteen minutes
in this alignment, the operators noticed that they had not seen
the expected temperature decrease in the system. Initially, the
control room operators dispatched an AO to increase the amount of
component cooling water being supplied to the RHR heat exchanger.
The operators still did not see the expected temperature decrease,
so they dispatched an AO to check the position of valve RHR-743
which was to have been providing the recirculation flowpath.
Recirculation flow through this path is not indicated in the
control room. Initially, the AO reported that the valve was open.
The control room operators then instructed the AO to verify flow
on local indicator FI-608. The AO reported that there was no flow
indicated. It was concluded that valve RHR-743 was closed. The
AO was instructed to open the valve. After the valve was opened,
the control room operators detected an immediate decrease in
temperature of the RHR system. By this time, the A RHR pump had
been run for approximately 66 minutes with little or no
appreciable flow.
Valve RHR-743 had been verified to be open on May 28, 1995, during
the performance of Operations Surveillance Test OST-163, Safety
Injection Test, and is required to be open as an Initial Condition
of GP-002. Ultimately, this mis-configuration resulted in the A
RHR pump being declared inoperable. This in turn forced the
6
licensee to return the unit to cold shutdown to facilitate the
disassembly and inspection of the pump.
The misalignment of valve RHR-743 constitutes one of six examples
which collectively constitute Violation 50-261/95-19-01,
Operations Configuration Control Events Concerning RHR Pump Flow
Path, SI-883R, Steam Driven Auxiliary Feedwater, And Containment
Ventilation Unit.
On June 9, 1995, after the A RHR pump had been inspected and
reassembled, control room operators were aligning the pump to
place it in service. At the time, the B RHR pump was supplying
decay heat removal in a configuration which bypassed its heat
exchanger. In this configuration, valve HCV-758, the common
discharge from both RHR trains' heat exchangers, was closed. The
operators started the A RHR pump, and stopped the B pump. They
immediately noticed that RHR flow decayed rapidly, restarted the B
pump and secured the A pump. Ultimately, the operators determined
that valve HCV-758 was closed and the A RHR pump had been started
without a flow path. The operators opened cross connect valve,
RHR-757C, restarted the A pump and successfully placed it in
service. The pump was operated for approximately two minutes with
only minimal flow afforded by the fact that valve HCV-758 leaked
by.
The inspectors concluded that procedure OP-201 was inadequate in
that it did not align the system to facilitate a flow path for the
A RHR pump before having the operator start it. This constitutes
one of six examples which collectively comprise Violation
50-261/95-19-01, Operations Configuration Control Events
Concerning RHR Pump Flow Path, SI-883R, Steam Driven Auxiliary
Feedwater, And Containment Ventilation Unit.
Reduced Inventory Operations
On June 9, 1995, the licensee initiated a draindown of the RCS in
accordance with GP-008, Draining The Reactor Coolant System, to
facilitate repairs of a leak on the main flange of RCP C. During
the repairs, RCS level was reduced to 43 inches below the main
vessel flange.
The inspectors reviewed the licensee's preparations for entry into
the reduced inventory condition on June 7, 1995. Licensee
preparations and precautions for a reduced inventory/mid-loop
operations were reviewed by the inspector. No deficiencies were
noted during this review. The inspectors witnessed portions of
the draindown on June 9; as well as, RCS level stabilization
immediately following draindown termination on June 10, 1995.
Additionally, the inspectors monitored operator performance during
routine control room tours while RCS inventory was below the
7
The inspectors concluded that appropriate sensitivity to risks
associated with operation in reduced inventory was displayed by
Operations personnel and the performance of operators during this
evolution was good.
AFW Pump Auto Start During Generator Draindown
At 5:33 p.m., on June 14, 1995, both MDAFW pumps started and the
SG blowdown isolation valves on all three SGs closed due to a low
low level in steam generator B. This occurred while draining the
steam generators in preparation for plant startup. In response to
this event, the operators defeated the AFW pump auto-start logic,
stopped the MDAFW pumps, and reopened the blowdown isolation
valves. AT 6:53 p.m. that day, the licensee made a 4-hour non
emergency report to the NRC in accordance with 10 CFR 50.72
(b)(2)(ii), ESF Actuation. A condition report was generated by
the licensee.
In response to this event, the inspectors reviewed Operating
Procedure OP-406, Steam Generator Blowdown/Wet Layup System;
Administrative Procedure AP-006, Procedure Use And Adherence; log
entries associated with the event; and Operations Management
Manual Procedure OMM-001, Conduct of Operations; reviewed the
auxiliary feedwater pump startup logic diagram, the ERFIS sequence
of events printout, the events notification worksheet, and
interviewed the AO and SRO involved in the event.
The inspectors determined that the event occurred as a result of a
failure by Operations personnel to appropriately block the SG low
and low-low level signals from the MDAFW autostart logic circuit
during the draindown of the generators. Blocking these inputs is
performed by repositioning 4 key switches in the back of the RTGB
from the "normal" to "defeat" position.
Draining of the steam generators was performed in accordance with
OP-406. This procedure requires that the 4 key switches be taken
to the "defeat" position prior to draining the generators. The
inspectors noted that although the AO initialed OP-406 as having
verified these key switches were positioned to the defeat
position, the switches were found in the "normal" position
following the event. The AO stated that while performing OP-406,
he called the SRO in the control room to request verification that
the 4 key switches were in the "defeat" position. Based on the
SRO's confirmation, the AO initialed the verification steps in the
procedure and continued.
The SRO advised the inspectors that his confirmation of the
defeated autostart circuit was based on noting the Train A and
Train B AFW Auto Initiation Defeated warning lights on the RTGB
were illuminated. This approach was flawed since these warning
lights can be illuminated without the 4 key switches specified in
OP-406 being in defeat position.
8
The inspectors concluded that the failure to adequately verify the
position of the 4 key switches prior to draining the steam
generators was contrary to the requirements of OP-406. This is
identified as one of six examples which collectively constitute
VIO 50-261/95-19-01, Operations Configuration Control Events
Concerning RHR Pump Flow Path, SI-883R, Steam Driven Auxiliary
Feedwater, And Containment Ventilation Unit.
Inadequate Containment Closeout
On June 3 and 4, 1995, the inspectors conducted inspections of
containment to verify the adequacy of the licensee's containment
closeout. This closeout was conducted in accordance with Plant
Program Procedure PLP-006, Containment Vessel Inspection/Closeout.
The areas toured by the inspectors included, but were not limited
to: all pump bays, the pressurizer cubicle, and the operating
deck. Numerous examples of loose tools, equipment and debris were
identified by the inspectors and reported to the licensee.
Additional cleanup of the CV was conducted by the licensee.
The plant startup was subsequently aborted and the RCS cooled down
to conduct repairs to RCP C. After the repairs to RCP C were
complete, the licensee commenced an RCS heatup in preparation for
reactor plant startup. Following the licensee's completion of
PLP-006, the inspectors again conducted a containment inspection
to verify the adequacy of the licensee's closeout. While the
general cleanliness had improved, the inspectors again found
numerous examples of loose equipment and debris. These were again
identified to the licensee for disposition.
Due to their size and weight, it is probable that many of the
items identified by the inspectors would not have been transported
to the ECCS sump during a LOCA. However, given the abundance, the
ease of detection, and prior inspector observations of deficient
CV closeout, the inspectors concluded that the licensee's efforts
at CV closeout were inadequate. This is identified as a weakness
in the licensee's containment closeout process.
Throughout the outage and following tours of containment, the
inspectors expressed concerns to licensee management regarding
loose paint in containment. Primarily, these concerns centered on
numerous areas of loose paint on the floor of the first level of
the CV, but the inspectors also noted areas of peeling or loose
paint on the operating deck, polar crane, and several of the HVH
units.
In response to these concerns, the licensee removed some of the
loose paint from the floors in containment and the HVH units and
provided the inspectors with documentation related to the generic
issue of loose paint in containment. This information did not
completely resolve the situation at H.B. Robinson. Pending
9
further review, this issue this is identified as an Unresolved
Item 50-261/95-19-03, Loose Paint In Containment.
b.
Followup - Operations (92901)
Inadequate Clearance For Work On Valve V1-8A
On April 17, 1995, routine preventive maintenance was to be
performed on valve V1-8A, one of three motor-operated valves which
supply motive steam to the SDAFW pump. Valve MS-20 which is
immediately upstream of V1-8A, was not closed. As a result, the
SDAFW pump started when valve V1-8A was opened.
At the end of report period for Inspection Report 95-14, the
inspectors had not completed their review of the circumstances
associated with this event. Accordingly, this issue was tracked
as Unresolved Item, 50-261/95-14-02, Inadequate Clearance For Work
On Valve V1-8A.
The inspectors reviewed the clearance, 95-00748, and reviewed
Operations Management Procedure OMM-005, Clearance And Test
Request. The clearance did not address valve MS-20. At the time
of the event, valve MS-20 was open. Accordingly, when valve V1-8A
was opened, steam was admitted to the SDAFW pump resulting in an
inadvertent start.
Procedure OMM-005, requires in part that all valves necessary to
protect personnel and equipment are properly closed or open as
necessary.
Clearance LCTR 95-00748 was inadequate in that it did not specify
a position for valve MS-20. Ultimately, this resulted in a
misconfiguration and inadvertent operation of the SDAFW pump.
It is noted that the planning of this work activity was inadequate
in that the maintenance on V1-8A did not adequately address the
operability of the SDAFW pump. When V1-8A was opened during the
event, and the "SDAFW Pump Low Discharge Pressure Trip"
annunciator was received, operations personnel questioned the
operability of the pump. Operations personnel appropriately
declared the pump inoperable and entered TS 3.4.4. until the
operability concern could be resolved.
The operability evaluation was performed by the system engineer.
Using the electrical logic and control wiring diagrams, the system
engineer concluded that the SDAFW pump would be inoperable if
V1-8A were greater than 96 percent open and the SDAFW pump had not
started, since valves V1-8B and V1-8C, the other two steam supply
valves to the SDAFW pump, would not open upon the receipt of a
valid start signal.
10
Historically, this preventative maintenance had been performed
with the unit in cold shutdown, this was the first time it had
been attempted with the unit on line. Although this activity had
been reviewed by operations and technical support personnel,
operability of the SDAFW pump had not been adequately evaluated.
During the event, annunciator APP-007-F5, "SDAFW Pump Low
Discharge Pressure Trip,"
was received. It is believed this
alarm may have been received during past performance of this
maintenance; however, operability of the pump was not questioned
at that time since the plant had been in cold shutdown during the
activity.
The technical review of this work activity was inadequate in that
the planned activity resulted in the misconfiguration and
inoperability of the SDAFW pump.
This issue constitutes one of six examples which collectively
comprise Violation 50-261/95-19-01, Operations Configuration
Control Events Concerning RHR Pump Flow Path, SI-883R, Steam
Driven Auxiliary Feedwater, And Containment Ventilation Unit.
Unresolved Item 50-261/95-14-02, Inadequate Clearance For Work On
Valve V1-8A is closed.
4.
MAINTENANCE
a.
Maintenance Observation (62703)
The inspectors observed safety-related maintenance activities on
systems and components to ascertain that these activities were
conducted in accordance with TS, approved procedures, and
appropriate industry codes and standards. The inspectors
determined that these activities did not violate LCOs and that
required redundant components were operable. The inspectors
verified that required administrative, material, testing,
radiological, and fire prevention controls were adhered to. In
particular, the inspectors observed/reviewed the following
maintenance activities detailed below:
WR/JO 94-AQYY1
Thermal Overload Testing (SI-860B)
WR/JO 95-AGGG1
Troubleshoot Cause Of Instrument Air
Compressor Breaker Fire
Flux Thimble Replacement
WR/JO 95-AHDB1
Troubleshoot RHR Pump Fails To Start
During OST-163
Upper Internals Installation
On May 22, 1995, the inspectors witnessed the installation of the
reactor vessel upper internals which was accomplished in
accordance with Maintenance Refueling Procedure MRP-005, Upper
Internals Removal and Installation. Overall, the internals lift
and installation were well conducted. However, the inspectors
noted that the subsequent lifting rig removal and return to the
storage stand were not as well orchestrated. During this phase of
the evolution, the inspectors observed the lifting rig impact the
manipulator crane, the wall of the refueling cavity, and an
electrical cord at the side of the cavity. None of these impacts
was particularly severe, but, this performance represented a
marked degradation below that observed by the inspectors for the
same basic activities only moments before. The inspectors
discussed these observations with the refueling coordinator and
were subsequently advised that a Condition Report would be
initiated to address this event.
FMEA Procedure Not Followed
On May 23, 1995, during a routine tour of containment, the
inspectors observed a worker in the reactor vessel head storage
area who was not logged into the area on the posted Foreign
Material Accountability Log Sheet. When questioned, the
individual acknowledged not logging into the area and attributed
his failure to not observing the warning sign posted at the
entrance to the FMEA area. The individual exited the area and a
condition report was generated. The inspectors were advised later
that the individual was counselled by licensee management on his
actions.
In response to this issue, the inspectors reviewed Plant Programs
Procedure PLP-047, Foreign Material Exclusion Area Program. The
inspectors also reviewed the condition report generated by the
licensee and interviewed the responsible supervisor. From this
review, the inspectors noted that PLP-047 established the head
storage area as a FMEA. As such, the individual was required to
log into the area and abide by other requirements to minimize the
potential of foreign material introduction into the reactor vessel
head. Overall, the inspectors concluded that the worker's failure
to log into the area was a violation of the requirements of
PLP-047. This failure constitutes a violation of minor
significance and is being treated as a non-cited violation,
consistent with Section VII of the NRC Enforcement Policy. This
is identified as NCV 50-261/95-19-04, FMEA Procedure Not Followed
In Head Storage Area.
Vessel Head Lift
On May 24, 1995, the inspectors witnessed a portion of the reactor
vessel head installation accomplished in accordance with
Maintenance Refueling Procedure, MRP-004, Reactor Vessel Head
Removal and Installation. This observation included head movement
from the storage stand to placement on the vessel.
The inspectors
12
also attended the pre-job brief. Overall, the conduct of the
evolution was good. Noteworthy strengths included lift team
coordination and communications. Strong management involvement
was also observed.
Inadvertent RHR Pump Start
On May 29, 1995, the inspectors witnessed portions of
troubleshooting performed to determine the cause of the B RHR pump
not starting during the performance of Operations Surveillance
Test OST-163, Safety Injection Test and Emergency Diesel Generator
Auto Start On Loss Of Power And Safety Injection And Emergency
Diesel Trips Defeat.
To facilitate troubleshooting, the RHR pump motor breaker was
racked to the test position. A defective relay was detected which
was removed, and taken to the I & C shop for further
troubleshooting. Subsequently, Operations racked-in the pump
motor breaker in the event RHR B pump was needed since the normal
pump starting circuitry was not affected by the aforementioned
relay. A member of the I & C troubleshooting team was informed of
the change in breaker position, but failed to advise the other
individuals involved in the repair effort.
Subsequently, a new relay was installed and when jumpers were
installed to verify its proper operation, the RHR pump motor B
started. Control room personnel immediately secured the pump.
Ultimately, the B RHR pump was successfully tested during a later
part of OST-163.
10 CFR 50, Appendix B, Criterion XIV requires that measures be
established for indicating the operating status of structures,
systems, and components, to prevent inadvertent operation. The
inspectors concluded that the licensee failed to establish
adequate measures to prevent the inadvertent start of the RHR
pump. This is contrary to the requirements of 10 CFR 50
Appendix B and is identified as a violation, VIO 50-261/95-19-05,
RHR Pump Start Due To Troubleshooting.
b.
Surveillance Observation (61726)
The inspectors observed certain safety-related surveillance
activities on systems and components to ascertain that these
activities were conducted in accordance with license requirements.
For the surveillance test procedures listed below, the inspectors
determined that precautions and LCOs were adhered to, the required
administrative approvals and tagouts were obtained prior to test
initiation, testing was accomplished by qualified personnel in
accordance with an approved test procedure, test instrumentation
was properly calibrated, the tests were completed at the required
frequency, and that the tests conformed to TS requirements. Upon
test completion, the inspectors verified the recorded test data
13
was complete, accurate, and met TS requirements, test
discrepancies were properly documented and rectified, and that the
systems were properly returned to service. Specifically, the
inspectors witnessed and/or reviewed portions of the following
test activities:
OST-163
Safety Injection Test and Emergency Diesel
Generator Auto Start On Loss Of Power And
Safety Injection And Emergency Diesel
Trips Defeat
Reactor Vessel Level Instrumentation
(System Calibration)
No violations or deviations were identified.
5.
ENGINEERING
Emergency Load Sequencing Timers (92903)
OST-163, Safety Injection Test And Emergency Diesel Generator Auto
Start On Loss Of Power And Safety Injection And Emergency Diesel
Trips Defeat, revision 24, included verification that emergency
loads sequenced onto the emergency buses at the appropriate times.
During two partial OST-163 performances on May 28, most of the
individual loads sequenced onto the emergency buses approximately
0.1 or 0.2 seconds outside the procedure's acceptance criteria.
Subsequent licensee investigation determined that the timing
relays had been improperly set earlier in that RFO.
The timers
were recalibrated and the applicable portion of OST-163 involving
the emergency bus load timing sequences was successfully completed
on May 29.
The inspectors reviewed the circumstances surrounding the improper
timer calibrations. Documents reviewed included: M-1035,
Emergency Load Sequencer Relay Replacement, and its field
revisions 1 and 4, that installed and initially calibrated the
digital timing relays; draft SP-1056, Time Delay Relay Calibration
Safeguards Train B, that was written but never issued to calibrate
the B train timing relays; Maintenance Procedure Revision/New
Procedure Request Form dated August 6, 1993, that requested
maintenance write calibration procedures for the timers; PIC-018
(020), Time Delay Relay Calibration Safeguards Train B (A), and
their associated document review packages and safety analyses; and
completed PIC-018 and 020 performed this RFO. In addition, the
inspectors interviewed cognizant maintenance and engineering
personnel who were either involved with the development of PIC-018
and 020 or participated in the investigation into the calibration
problem. The system engineer who developed the draft SPs and
interfaced with maintenance during M-1035 implementation and the
14
development of PIC-018 and 020 had retired from the company. The
inspectors confirmed that the licensee's investigation had
identified the contributing causes that resulted in the timers
being improperly calibrated during that RFO.
CR No. 95-01379, approved June 7, 1995, documented the causes and
proposed corrective actions to address the improper timer
calibrations. The primary cause was personnel error that resulted
in a failure to ensure design values developed for field
Revision 4 to M-1035 were properly transferred to design
documents.
For example, drawing 5379-3238 was not revised to
reflect that the actual timer set points were to be adjusted for
the times required for the logic circuits to actuate and close
their associated load breakers. A planned corrective action
identified in CR 95-01379 was to provide lessons learned from the
event to the engineering staff. Also, optimum timer settings were
to be established and associated maintenance procedures revised
accordingly. The inspectors considered that these actions were
appropriate to preclude recurrence of this event.
The failure to incorporate design information into appropriate
design documentation such that sequencing timer calibration
procedures were established with improper values was a violation
of 10 CFR 50, Appendix B, Criterion III.
The violation has
minimal safety significance, in that, the amount the timers were
outside the expected values was not sufficient to adversely
affect emergency bus loadings and the unit was never operated with
the improper settings. This licensee identified and corrected
violation is being treated as a non-cited violation consistent
with Section VII of the NRC Enforcement Policy. Thus, this item
is identified as NCV 50-261/95-19-06, Failure To Incorporate
Sequencing Timer Settings Into Appropriate Design Documents.
During the event review, the inspectors identified that during the
second test performed on May 28, the A AFW pump breaker closed
approximately 0.8 seconds later than the B AFW pump breaker.
Review of the previous three tests (two in 1993) and the
subsequent successful test on May 29, revealed that the A AFW pump
breaker typically closed within 0.1 seconds of the B AFW pump
breaker. Maintenance personnel indicated that they had taken no
action to review the occurrence since Operations had not informed
them of the abnormal reading. Further review revealed that
personnel performing the test and reviewing the test results had
failed to identify this discrepancy. Not performing a
sufficiently detailed review of OST-163 test data to identify a
potentially intermittent condition for additional review or future
monitoring was considered a weakness.
During this inspection, the inspectors noted that the calibration
frequency for the sequencing timers was every third RFO.
Comparison of the as-left timer settings from M-1035 field
revision 4 and the as-found values per calibration procedures
15
PIC-018 and 020 indicated that the timers had drifted a maximum of
0.8% (0.04 seconds) between December 1989 and May 1995. Thus, the
inspectors concluded that an every third RFO calibration interval
was acceptable.
Electrical Maintenance and Modifications (62705)
During the refueling outage, the licensee implemented
modifications to the main control room aimed at improving the
layout from the human factors viewpoint. The modification
consisted of removing a fairly large control panel, which was
greatly under utilized as a result of previous modifications.
Removal of the panel created additional space in the central area
of the control room. Operator work stations were relocated and
upgraded thus achieving improved use of space in the control room.
Plant related non-safety annunciators which had been on the
deleted control panel were relocated within the control room, and
non-safety 230 Kv breaker status lamps were replaced by ERFIS data
points. A number of safety-related cables which had been routed
through the deleted control panel were removed and rerouted using
new cable. The modification was implemented under Engineering
Services Request (ESR) No.94-882.
Due to extensive wiring changes taking place in a relatively short
time period, the NRC inspected the controls that the licensee
employed to ensure that the changes were correctly implemented.
Requirements relevant to the area of inspection were 10 CFR 50.59,
Changes, Tests and Experiments, and 10 CFR 50, Appendix B,
Criterion III, Design Control.
The inspection focused on wiring changes. Specific inspection
activities and findings were as follows:
Walkdown inspection of the equipment, raceways and cables in
the main control room, relay panels/spreading room and
control room roof involved with ESR 94-882. The inspector
concluded that the work was done according to the licensee's
installation specification, including conduit fill and
pulling points, and the quality of workmanship was good.
In relation to safety-related cables C21732C and C21732D,
which were multi-conductor cables selected at random, the
inspector verified the following attributes: wires were
landed on the correct relay panel terminals, correct size
lugs were used, and correct size crimping tool was used. In
addition, the inspector verified calibration of the crimping
tool.
The inspector reviewed the completed post-modification test
sheets for the modified circuits, and verified that the
testing was adequate and results good.
16
The inspector was told that annunciator points associated with
switchyard equipment were deleted from the control room by ESR 94-882. These points were:
OCB 52-8 [generator breaker] failure detection trouble
North 230 kV bus breaker failure lockout
North 230 kV bus differential lockout
OCB 52-9 [generator breaker] failure detection trouble
It could not be determined during the inspection what the original
basis was for having these particular four points in the main
control room. The inspector inquired as to whether important
information was lost as a result of deletion of these annunciator
points. The system engineer assigned to coordinate with the
Transmission Department stated that these four points were
repeated on an annunciator in the switchyard relay house. He also
stated the annunciator points were repeated at the transmission
system control center in Raleigh, N. C., and that the dispatcher
would notify the nuclear plant control room operator should the
annunciator go to alarm condition. The inspector indicated to the
licensee that he wanted to verify the annunciators in the
switchyard relay house. This activity was scheduled for the
following day. The following day the system engineer stated that
the two generator breaker failure detection trouble annunciators
were not at the annunciator panel in the switchyard relay house
and therefore were not repeated at the Raleigh center. Instead,
the breaker failure relays were monitored by lamps, which were
mounted on the front of the respective breaker control panels. A
supervisory lamp is considerably different than an annunciator
because an annunciator gives immediate information to system
operators whereas a lamp can only give information when operators
visit the relay house, which was reported to be about once per
month. The inspector went to the relay house and verified the
annunciator inscriptions and the breaker failure supervisory
lamps.
The safety evaluation for ESR 94-882 indicated the following: "The
switchyard annunciator APP-033 alarm lights on the 230 kV Line
Panel are removed. Alarms for four of the lights are repeated on
an annunciator in the switchyard building. The activity maintains
the alarm functions associated with APP-033." The design basis
document for ESR 94-882 indicated that: "The annunciator lights
[from APP-033] are not required since their functions are repeated
elsewhere; therefore, these lights will be deleted from the
control room."
The inspector noted that the safety evaluation and the design
basis document did not accurately describe the change because, as
stated above, two of the annunciators in question were in fact not
repeated elsewhere. The fact that these documents were not
accurate in this regard was considered significant by the
inspector. As far as could be determined through discussions with
17
licensee personnel, persons preparing the safety evaluation
misinterpreted statements made by transmission system engineers as
to the design of the annunciators at the relay house. More
significantly, apparently no attempt was made to verify the
particular description in the safety evaluation. The inspector
concluded that the safety evaluation could be revised to support
deletion of the two non-safety-related annunciators in question.
The fact that the original information was not correct represents
a weakness in the sense that, should the licensee continue to
allow unverified statements to form the basis for conclusions in
their safety evaluations, inadequate safety evaluations could
result.
Overall, the inspector concluded that the wiring changes to the
main control room performed under ESR 94-882 were well
implemented. This conclusion was based on results of walkdown
inspections, detailed verification of representative cables, the
post-modification test results, and discussions with engineers.
6.
EXIT INTERVIEW
The inspectors met with licensee representatives (denoted in paragraph
1) at the conclusion of the inspection on June 28, 1995. During this
meeting, the inspectors summarized the scope and findings of the
inspection as they are detailed in this report. The licensee
representatives acknowledged the inspector's comments and did not
identify as proprietary any of the materials provided to or reviewed by
the inspectors during this inspection. No dissenting comments from the
licensee were received.
Item Number
Status
Description
URI 95-14-02
Closed
Inadequate Clearance For Work On
Valve V1-8A.
VIO 95-19-01
Opened
Operations Configuration Control
Events Concerning RHR Pump Flow
Path, SI-883R, Steam Driven
Auxiliary Feedwater, And Containment
Ventilation.
URI 95-19-02
Opened
SI Pump Breaker Racked-In With LTOPP
In Service.
URI 95-19-03
Opened
Loose Paint In Containment.
NCV 95-19-04
Open/Closed
FMEA Procedure Not Followed In Head
Storage Area.
VIO 95-19-05
Opened
RHR Pump Start Due To
Troubleshooting.
Item Number
Status
Description
NCV 95-19-06
Opened/Closed
Failure To Incorporate Sequencing
Timer Settings Into Appropriate
Design Documents.
7.
ACRONYMS AND INITIALISMS
Auxiliary Operator
CFR
Code Of Federal Regulation
CIT
Clearance Information Tag
CR
Control Room, Condition Report
CV
Containment Vessel
ERFIS
Emergency Response Facility Information System
Engineered Safety Feature
Foreign Material Exclusion Area
HVH
Heating Ventilation Handling
Instrumentation And Control
LCO
Limiting Condition for Operation
Loss Of Coolant Accident
LTOPP
Low Temperature Over Pressure Protection
Motor Driven Auxiliary Feedwater
OMM
Operations Management Manual
PLP
Plant Program Procedure
Power Operated Relief Valve
Reactor Cooling Pump
Steam Driven Auxiliary Feedwater
Safety Injection
Senior Reactor Operator
TS
Technical Specification