ML14178A401
| ML14178A401 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 11/08/1993 |
| From: | Christensen H, William Orders NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A399 | List: |
| References | |
| 50-261-93-21, NUDOCS 9311160127 | |
| Download: ML14178A401 (20) | |
See also: IR 05000261/1993021
Text
NR REG4
UNITED STATES
Co
NUCLEAR REGULATORY COMMISSION
REGION II
o
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report No.:
50-261/93-21
Licensee:
Carolina Power and Light Company
P. 0. Box 1551
Raleigh, NC
27602
Docket No.:
50-261
License No.: DPR-23
Facility Name: H. B. Robinson Unit 2
Inspection Conducted: September 12 - October 16, 1993
Lead Inspector:
-
t.sn
P-'-
///5 95$
W. T. Orders, Seni
Resident Inspector
Date Signed
Other Inspector:
C. R. Ogle, Resident Inspector
C. W. Rapp, Regional Inspector
N. Merriweather, Regional Inspector
Approved by:
_
7
_ _ __ _
_________
_73
)' 0. C ristensen, Chief
Date Signed
Reactor Projects Section lA
Division of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection was conducted in the areas of operational
safety verification, maintenance observation, followup of previously
identified items, and routine followup
Results:
One violation was identified which involved two examples of personnel failing
to follow procedures. The issues concerned an operator opening the incorrect
electrical breaker, and the failure to control a sludge lance vent path.
(paragraph 3)
A second violation was identified which concerned a failure to establish
containment integrity before commencing refueling operations. (paragraph 3)
Two non cited violations were identified involving an STA who departed the
site before the end of his shift and an area fire watch who vacated his post
before being relieved of his duties. (paragraph 3)
9311160127 931108
PDR ADOCK 05000261
G
2
Two inspector followup items were identified involving a loose part in a fuel
assembly, and fractures in the terminal posts of the vital batteries.
(paragraphs 3 and 4)
REPORT DETAILS
1.
Persons Contacted
R. Barnett, Manager, Project Management
J. Benjamin, Shift Outage Manager, Outages and Modifications
S. Billings, Technical Aide, Regulatory Compliance
B. Clark, Manager, Maintenance
- T. Cleary, Manager, Technical Support
- D. Crook, Senior Specialist, Regulatory Compliance
- C. Dietz, Vice President, Robinson Nuclear Project
R. Downey, Shift Supervisor, Operations
J. Eaddy, Manager, Environmental and Radiation Support
G. Elam, Program Manager, EGS Corporation
S. Farmer, Manager, Engineering Programs, Technical Support
R. Femal, Shift Supervisor, Operations
- W. Flanagan Jr., Operations Manager
R. Hardy, Test Director, Wyle Laboratories
B. Harward, Manager, Engineering Site Support, Nuclear Engineering
Department
P. Jenny, Manager, Emergency Preparedness
D. Knight, Shift Supervisor, Operations
D. Labelle, Project Engineer, Nuclear Assessment Department Site Unit
A. McCauley, Manager, Electrical Systems, Technical Support
R. Moore, Shift Supervisor, Operations
D. Morrison, Shift Supervisor, Operations
- T. Niemi, Project Engineer, Nuclear Assessment Department
D. Nelson, Shift Outage Manager, Outages and Modifications
A. Padgett, Manager, Environmental and Radiation Control
- M. Pearson, Plant General Manager
D. Seagle, Shift Supervisor, Operations
M. Scott, Manager, Reactor Systems, Technical Support
- E. Shoemaker, Manager, Mechanical Systems, Technical Support
W. Stover, Shift Supervisor, Operations
- A. Wallace, Manager, Shift Operations, Operations
D. Waters, Manager Regulatory Affairs
- D. Whitehead, Manager, Plant Support Services
D. Winters, Shift Supervisor, Operations
P. Yandow, EQ Coordinator, CP&L
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
NRC Managements Visits
H. Christensen, Chief, Projects Section lA, Division of Reactor
Projects, visited the site on October 15, 1993. Mr. Christensen toured
the facility with the residents and attended the exit meeting on
October 15, 1993. He also met with members of the licensee's management
organization.
- Attended exit interview on October 15, 1993.
2
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Plant Status
Refueling Outage 15 continued during the report period with fuel
assembly reload ongoing at the end of the inspection period. The outage
is currently scheduled to end on November 3, 1993.
3. Operational Safety Verification (71707)
The inspectors evaluated licensee activities to confirm that the
facility was being operated safely and in conformance with regulatory
requirements. These activities were confirmed by direct observation,
facility tours, interviews and discussions with licensee personnel and
management, verification of safety system status, and review of facility
records.
To verify equipment operability and compliance with TS, the inspectors
reviewed shift logs, Operation's records, data sheets, instrument
traces, and records of equipment malfunctions. Through work
observations and discussions with Operations staff members, the
inspectors verified the staff was knowledgeable of plant conditions,
responded properly to alarms, adhered to procedures and applicable
administrative controls, cognizant of in-progress surveillance and
maintenance activities, and aware of inoperable equipment status. The
inspectors performed channel verifications and reviewed component status
and safety-related parameters to verify conformance with TS. Shift
changes were routinely observed, verifying that system status continuity
was maintained and that proper control room staffing existed. Access to
the control room was controlled and operations personnel carried out
their assigned duties in an effective manner. Control room demeanor and
communications were appropriate.
Plant tours and perimeter walkdowns were conducted to verify 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.
Shift Technical Advisor (STA) Departs Before End Of Shift
On the morning of September 16, 1993, the inspectors were present in the
control room during turnover of the oncoming day shift, when it was
noted that the off-going STA was absent. From a subsequent review of
security access records, the inspector determined that the offgoing STA
had left the protected area approximately 5 minutes prior to the arrival
of the oncoming STA.
Although the inspectors noted from a review of TIS 6.2.3.c, that the STA
was not required during cold shutdown conditions, an individual was
specified to fill the STA position by the watchstanders listing posted
3
in the control room. Operations Management Manual, OMM-008, Minimum
Equipment List and Shift Relief, discusses turnover responsibilities for
the STA position. Though some of the responsibilities of OMM-008 are
not applicable while in cold shutdown, some are meaningful during a
turnover in any plant condition. Additionally, OMM-008 specifically
requires that shift operating personnel remain on duty with full
responsibilities of their position until properly relieved. OMM-008
requires that this turnover be conducted at the normal watchstation.
The inspectors requested that the licensee provide details concerning
the premature departure of the STA. Licensee management advised the
inspectors that the individual left without having been granted
permission by the onwatch shift supervisor and that he had done so of
his own volition because he felt that he was not needed due to plant
conditions.
The licensee counselled the individual concerning management's
expectation that watchstanders remain until properly relieved. The
licensee also stated their intention to review this expectation with all
watchstanders.
The failure of the STA to conduct a watch turnover is a violation of the
requirements of OMM-008. However, this identified violation is not
being cited because criteria specified in Section VII.B of the NRC
Enforcement Policy were satisfied. This is identified as a non-cited
violation, NCV 93-21-01: Failure Of STA To Conduct Turnover Required By
OMM-008.
Sewage Overflow In Protected Area
At approximately 8:00 a.m. on September 17, 1993, the licensee
discovered sewage overflowing a manhole in the protected area. The
licensee stopped the overflow and a septic tank service was utilized to
lower the system level.
In response to this overflow, the licensee notified the South Carolina
Department of Health and Environmental Control on September 17, 1993.
As a result of the notification to the state, the licensee made a 4-hour
non-emergency notification to the NRC in accordance with the
requirements of 10 CFR 50.72 (b) (2) VI.
The licensee notified the
resident inspectors prior to the 10 CFR 50.72 notification.
On October 7, 1993, two subsequent sewage spills occurred at the site.
These spills were stopped and recovery actions were taken by the
licensee. As a result of notification of state authorities, the
licensee made initial and followup notifications to the NRC in
accordance with 10 CFR 50.72 (b) (2) VI at 2:25 p.m. and at 5:30 p.m. on
October 7, 1993.
4
Based on their review of these events, the inspectors concluded that the
licensee met the requirements for NRC notification specified in 10 CFR
50.72. The inspectors have no further questions on these events.
Reactor Coolant System Draindown
On September 17, 1993, the inspectors witnessed the performance of
General Procedure, GP-008, Draining The Reactor Coolant System. This
draindown was accomplished to reduce RCS level to approximately 10
inches below the reactor vessel flange to support vessel head removal.
The inspection effort included a partial walkdown of the reactor coolant
level standpipe system prior to the draindown, attending the pre-shift
briefing for the evolution, and observation of control room activities
during the level change.
Overall, the inspectors concluded that the evolution was well controlled
with a proper emphasis on safety. There was a strong effort by
operations personnel to ensure that RCS level was closely monitored and
that appropriate overlap existed in available level instruments. The
Management Designated Monitor conducted the pre-shift briefing which
included information from previous draindown events. He also remained
present in the control room during the level change. The inspectors
have no further questions on this evolution.
Operator Inadvertently Deenergized Motor Control Center MCC-5
At approximately 10:00 p.m. on September 19, 1993, power to MCC-5 was
interrupted when an operator inadvertently opened the normal power
supply breaker for the MCC. The operator had been dispatched to
transfer service water pump D to the emergency power supply, but in
fact, commenced the transfer sequence at the wrong switchboard. After
power to MCC-5 was interrupted, the operator recognized the mistake,
reclosed the normal power supply breaker, and informed the control room
of his actions. The MCC-5 loads which were stopped as a result of the
power interruption were then restarted. Service water pump D was
subsequently transferred to its alternate power supply.
The inspectors interviewed the operator involved; reviewed log entries
in the shift supervisor's and control operator's logs; and reviewed an
Off Normal Condition Analysis (ONCA) form generated for the event.
Based on this effort, the inspectors determined that the operator became
distracted while enroute to accomplish the transfer of service water
pump D to the alternate power supply. This distraction occurred while
the operator explained the operation of the plant's interlocked power
supply breakers to a watchstander trainee. As a result, the operator
positioned himself at the MCC-5 normal and emergency supply breakers as
opposed to the normal and emergency supply breakers for service water
pump D. These switchboards are similar in appearance and function, and
though not adjacent, are in close proximity to one another.
5
Technical Specification 6.5.1.1, Procedures, Tests, and Experiments
requires, in part, that written procedures be established, implemented,
and maintained covering the activities recommended in Appendix A of
Regulatory Guide 1.33, Rev. 2. 1978. Appendix A, Paragraph 4 requires
instructions for operation of onsite electrical systems. Operating
Procedure, OP-603, Electrical Distribution is provided these
instructions.
Contrary to the above, on September 19, 1993, an operator deviated from
OP-603 while transferring service water pump D from normal to emergency
power. As a result of this deviation, power was lost to motor control
center, MCC-5. This is the first of two examples which in the aggregate
comprise a violation, VIO 93-21-02: Failure To Follow Procedures, Two
Examples.
Sludge Lance Rig Vent Valve Misaligned
During the ongoing refueling outage the licensee accomplished sludge
lancing on the steam generators secondary sides. The equipment used to
accomplish this work was mounted on a tractor trailer external to the
containment. Two containment vent valves, V12-12 and V12-13, were
removed and necessary hoses and electrical cables to support the
evolution were routed through the idle penetration. To permit lancing
while containment integrity or containment closure were required, the
system incorporated a foam-filled CV penetration collar which bolted
onto the V12-12 flange, a "closed" sludge lance system outside of
containment and a vent hose from the sludge lancing equipment back into
containment. Each of the hoses which penetrated containment were also
equipped with a pair of isolation valves on either side of the
containment penetration. Special Procedure, SP-1231, S/G Sludge Lance
and Inspection was developed to provide the guidance necessary to
accomplish the lancing.
On September 21, 1993, during a routine tour, the inspectors observed
that a vent valve for the slurry tank on the trailer mounted rig was
open. At the inspector's request, the valve was confirmed to be open by
a contractor assigned to operate the sludge lancing rig. The valve was
then closed by the contractor.
This open valve on the slurry tank provided a direct vent path from the
rig to atmosphere and hence, since sludge lancing was in progress, a
direct path from the containment to atmosphere. The contractor
indicated that he was unsure why the valve was open. CV closure was in
effect at the time of this observation, however, CV integrity was not
required.
SP-1231 required that all vent paths on the sludge lance rig be
identified by the system engineer and caution tagged prior to opening
the isolation valves for the rig. The open vent valve identified by the
inspectors was not caution tagged. When questioned on the lack of a
caution tag, the cognizant supervisor stated that a walkdown had been
performed but failed to identity this valve as a potential vent path.
6
Technical Specification 6.5.1.1. Procedures, Tests, and Experiments,
requires, in part, that written procedures be established, implemented,
and maintained, for the activities recommended in Appendix A of
Regulatory Guide 1.33, Rev 2. 1978. Paragraph 9 of Appendix A requires
that maintenance be performed in accordance with written procedures.
Special Procedure, SP-1231, Steam Generator Sludge Lance and Inspection,
was written to provide instructions for conducting sludge lancing of the
steam generators secondary sides.
Contrary to the above, on September 21, 1993, a vent path to atmosphere
was not identified and caution tagged as required by SP-1231. As a
result, a vent valve on the sludge lancing rig was open and a vent path
from the containment existed. This is the second of two examples which
in the aggregate comprise a violation, VIO 93-21-02: Failure To Follow
Procedures, Two Examples.
The inspectors also noted that there was no procedural requirement in
SP-1231 to conduct another verification that the potential vent paths in
the sludge rig were properly isolated prior to establishing integrity
for refueling. When questioned by the inspectors, the system engineer
stated his intentions were to perform an additional system walkdown of
the sludge lancing rig prior to establishing integrity.
Area Fire Watch Vacates Post
On September 22, 1993, the licensee determined that an area firewatch
had vacated his assigned post prematurely. The watch had been stationed
in the Emergency Switchgear/Safeguards Room as required by Fire
Protection Procedure (FP-012), Fire Protection Systems Minimum Equipment
and Compensatory Actions, when the halon suppression system for that
area (zone 20) was disabled. The system was disabled at 8:34 p.m. on
September 22, 1993, to support maintenance in the room. At 11:50 p.m.
Operations personnel noted that the area fire watch had left the room
while zone 20 was still disabled.
The inspectors reviewed the shift supervisor's log and the on-shift fire
technician's log. Additionally, the inspectors interviewed the
cognizant maintenance and fire protection supervisors. Based on the
information obtained from this effort, the inspectors concluded that the
area fire watch was unaware of his responsibility to remain at his
assigned station until the fire suppression system was restored to
service.
In response to this event, the licensee committed to providing training
to area fire watches reiterating the need for the watch to remain posted
until the fire suppression system is fully returned to service. The
licensee has also developed a sheet which outlines the responsibilities
of area fire watches. As described by the licensee, this information
sheet will be provided to area fire watches upon deactivation of fire
zones. The inspectors reviewed the sheet and noted that it specifically
addressed the responsibility of area fire watches to remain in assigned
zones until authorized to depart by the on-shift fire technician.
7
The failure of the area fire watch to remain in Zone 20 is a violation
of the requirements of FP-012. This violation will not be subject to
enforcement action because the licensee's efforts in identifying and
correcting the violation meet the criteria specified in Section VII.B of
the Enforcement Policy. This is identified as a non-cited violation,
NCV 93-21-03: Area Fire Watch Vacates Post.
Failure To Maintain Containment Integrity During Fuel Movement
On September 27, 1993, the unit was in cold shutdown condition, in a
scheduled refueling outage with core off-load underway. At approximately
10:30 a.m. that morning, licensee Technical Support personnel were
requested to check the adequacy of the containment building equipment
hatch seals after the Resident Inspectors noted that the hatch was
leaking air into the building. The licensee visually verified air
inleakage, and at approximately 11:30 a.m., Operations personnel were
notified of the existence of the leakage. Although core off-load had
been completed approximately 10 minutes earlier, the leak path existed
during fuel movement operations.
The NRC operations duty officer was notified of this event at 3:21 P.M.
that afternoon as a condition that alone could have prevented the
fulfillment of the safety function of the structures or systems needed
to control the release of radioactive material.
Upon discovery of this event, attempts were made to eliminate the
leakage by tightening the hatch bolts, but these attempts were un
successful.
Subsequently, the equipment hatch was removed and the seals
and flange surfaces were inspected. The results of the inspection
revealed that the seals and seating surfaces were in relatively good
condition with one of the seals intact, and only minor damage noted on
the other seal.
The cause of this event is attributed to the lack of controls for the
re-installation of the equipment hatch during an outage, to support
activities requiring containment integrity, such as fuel movement.
Operations management manual procedure OMM-033 provides only limited
guidance for the re-installation of the hatch for containment
closure/refueling integrity. Adverse Condition Report 93-173 was
initiated to document this condition and to facilitate a root cause
investigation.
Plant Procedure CM-603 is used to secure the equipment hatch to meet
containment integrity requirements for reactor operations. Procedure
OMM-033, was developed to provide guidance for equipment hatch
installation to support, in part, refueling operation. However, this
procedure appeared to be inadequate to ensure that containment closure
(integrity) was achieved. In addition, the procedure was not used on
September 23, 1993, when the hatch was installed to support refueling
operations.
8
Technical Specification 3.8.1., refueling operations, requires that the
equipment hatch be "properly closed".
Implicit in this requirement is
the requisite that the hatch be capable of performing its intended
safety function, which in this case is to prevent the release of
radioactive material to the environment given a fuel handling accident
or a prolonged loss of core cooling.
Furthermore, Technical Specification 6.5.1.1, Procedures, Tests, and
Experiments requires, in part, that written procedures be established,
implemented, and maintained covering the activities recommended in
Appendix A of Regulatory Guide 1.33, Rev. 2. 1978. Appendix A,
Section 3.f(1) requires instructions for maintaining containment
integrity. Implicit in this requirement are the requisites that the
procedures be adequate to facilitate the applicable evolution and that
personnel use the procedures during the performance of the evolution.
Contrary to those requirements, on September 23, 1993, the licensee did
not use a procedure to facilitate the re-installation of the containment
building equipment hatch when they were preparing for refueling
operations. This resulted in the inadequate installation of the
equipment hatch in that it was not "properly closed" and as such was not
capable of performing its intended safety function. This is a Violation,
93-21-04: Failure To Maintain Containment Integrity During Fuel
Movement.
A review of this event revealed that there was only minor safety
significance involved with the leak at the equipment hatch. The major
vulnerability involves fuel damage during refueling operations which
could lead to the release of fission products to the containment
atmosphere. However, there does not appear to be a viable scenario
which would result in containment pressurization which in turn would
result in release of radioactive material to the environment.
Loose Part In Fuel Assembly
During the ongoing refueling outage, the licensee examined a number of
fuel assemblies for damage. To facilitate this inspection, the fuel
assembly tie plates were replaced with a special vendor supplied guide
plate. This guide plate was fastened to the control rod guide tubes
with three anchors. The anchors consist of a shaft and nut which
interface to expand a split tube. This expanded split tube provides a
grip with the guide tube inner diameter to hold the guide plate in
place. When the guide plate was removed from assembly U24, the shaft of
one of the three anchors was broken. The retaining nut, the split tube,
and the end of the shaft were missing from the tool.
At that time, the
vendor technicians performing this service, assumed that the pieces had
fallen into the spent fuel pit, but failed to notify the licensee.
Following the inspection of assembly U24, an attempt was made to load an
RCCA in the assembly. The vertical travel of the RCCA was blocked
approximately 2 feet above its fully loaded position. When the RCCA
would not insert into the assembly, it was inspected with a camera. The
9
split tube from the aforementioned fuel inspection tool was found to be
on the end of one of the RCCA fingers and consequently removed. Thus,
the remaining pieces in the guide tube were the retaining nut and the
shaft end. During a later inspection, it was verified that the
retaining nut is lodged at the transition area. Because of the design
of the tool, the shaft end is free of the retaining nut and is located
in the bottom of the guide tube, below the retaining nut.
The licensee elected to leave the tool fragments in the guide tube and
move the assembly to a non-controlled (no control rod) position in the
core. This decision was justified by the licensee's fuel vendor,
Siemens.
A copy of that evaluation was forwarded to the RII office for followup.
Pending completion of that analysis, this issue will be tracked as IFI
93-21-05: Fuel Assembly Loose Part.
ECCS Piping Flush
On September 27, 1993, the inspectors witnessed accomplishment of
Special Procedure, SP-1163: SI-891 C, SI-891 D, and SI-863 B Flush.
This special procedure was accomplished to flush selected ECCS piping to
the RWST to ensure that no white plastic remained in any uninspected or
unflushed portion of the system. Additional information concerning the
discovery and removal of the white plastic material is contained in NRC
Inspection Report 50-261/92-24. This procedure was accomplished
coincident with cavity draindown and directed flow to the RWST through
the flushed piping.
Overall, the accomplishment of SP-1163 was satisfactory. However, the
inspectors noted that total flowrate through the operating RHR pump
exceeded the maximum flow rate specified in Operating Procedure, OP-201,
Residual Heat Removal System. This was identified to the system
engineer and cognizant engineering supervisor. The inspectors requested
that the licensee review this issue to determine if the maximum runout
flow for the RHR pump had been exceeded. After their review, the
licensee provided documentation that despite exceeding the precautional
limit of OP-201, the pump flow had not exceeded runout conditions. The
inspectors independently reviewed the documentation and concurred with
the licensee's conclusions. The inspectors had no further questions on
Summary
Two separate noncited violations for watchstanders (an STA and a fire
watch) leaving their assigned duty prematurely indicates that
improvements in this area are warranted. The RCS partial draindown
activity was well controlled and received adequate monument attention.
A violation with two examples of failure to follow procedural
requirements is a concern because they represent a continuing pattern of
plant personnel not complying with procedures. A second violation was
10
identified for accomplishing a maintenance task without using the
applicable procedure. An inspector followup item was identified for
loose parts in a fuel assembly guide tube.
4. 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, and approved procedures. 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, and radiological controls were
adhered to. In particular, the inspectors observed/reviewed the
following maintenance activities:
WR/JO
92ARSN4
Discharge Testing of Battery
Charger Al
WR/JO
93AHYLl
Replace Flow Switch For HVE-16
GP-010
Refueling (Fuel Offload)
Retest Program For Environmental Qualification Of The Patel Conduit
Seals
On August 2, 1993, the NRC conducted a special inspection at Wyle Labs
in Huntsville Alabama to review the retest program for the EGS (formerly
Patel) conduit seals. This program was implemented by the licensee to
resolve 10 CFR 50.49 Environmental Qualification (EQ) issues identified
by the NRC in Inspection Report 50-261/91-03. The inspector met with
representatives from CP&L, EGS, and Wyle Labs to discuss the preliminary
results of the retest program. The licensee stated during this meeting
that the four test specimens exposed to the H.B. Robinson LOCA/
Submergence accident profiles did not experience any leakage past the
seals during the entire test. One anomaly was identified during the
test and it was resolved by extending the duration of the test. The
test specimens were examined prior to being removed from the LOCA
chamber. The test specimens/setup was observed to be similar to that
described in, Re-Test Procedure for Submergence Qualification of Conduit
Seals, (Report No. EGS-TR-841215-07, Revision B), however, unlike the
procedure end caps were added to the specimens to protect the wires
against direct spray.
The test specimens were removed from the LOCA
chamber and were inspected for visual signs of leakage past the seals
(e.g. grommet and wire integrity).
In addition the break away torque
was measured for one of the samples. The above inspections had
satisfactory results. The inspector examined the data sheets for the
last six days of the test which covered the period July 27 - August 1,
1993, with no anomalies observed. The inspection concluded with the
specimens being packaged for transporting to EGS for further examination
and study. The licensee indicated that the final test report should be
issued by the end of the year. This issue will remain open pending
review of the final test report and the revised EQ files.
Discharge Testing Of Battery Charger A-i
On September 21, 1993, the inspectors witnessed the setup and initial
steps accomplished to perform discharge testing of battery charger A-1.
The discharge test was aborted when the system engineer recognized that
the load bank had been incorrectly attached to charger A-1.
During installation of the cables to the load bank, the output power
leads to the output terminal block in the charger were removed. The
load bank cables were then attached to this terminal block. The intent
of the procedure was to remove the cables between the terminal block and
the DC bus and attach the load bank cables in their place.
Following the recognition of this deficiency, the load bank cables were
removed, the charger restored to normal configuration, and the test
rescheduled for a later date.
RWST Inspection
On September 30, 1993, the inspectors witnessed preparations for and the
subsequent entry of a diver into the RWST. This entry was accomplished
to permit an inspection of the RWST following a flush of selected ECCS
piping to the RWST in accordance with Special Procedure, SP-1163, SI
891C, SI-891D, and SI-863B flush. The diver's efforts were tracked by
personnel outside the tank on a monitor fed from a camera operated by
the diver.
As described by the diver, visibility in the tank was good with only a
few pieces of debris observed on the bottom. A few flakes of white
material were reported on the tank floor, however, they could not be
recovered due to their small size.
No discrete pieces of plastic were
reported by the diver (see paragraph 3 for additional information on the
plastic material).
Items recovered included small pieces of string, a
short piece of wire, and a bolt.
The overall performance of the evolution was satisfactory. However, the
inspectors noted that the benefit provided by monitoring the screen
outside the tank was marginal.
Both poor lighting used with the camera
and the small monitor size made vidio evaluation of objects by support
personnel difficult. This was a marked reduction in capability over
that observed by the inspectors during previous dives into the RWST.
The inspectors were informed that this particular camera/monitor setup
was used for this inspection due to the desired camera/monitor rig being
unavailable. The inspectors were also advised that the camera output
was being videotaped for potential future review. The inspectors had no
further questions on this evolution.
Use of Temporary Leak Sealants
The inspectors conducted a review of the controls associated with the
use of temporary leak sealants. This review included procedures,
management oversight, engineering and safety evaluations, and
12
application of temporary leak sealants on plant equipment. The
inspectors found that temporary leak sealants are used on both safety
and non-safety related equipment. Application of temporary leak
sealants was controlled by the temporary modification process. The
vendors procedures for installing the temporary leak sealant were
included in the temporary modification package and reviewed with the
temporary modification. The temporary modification also delineated the
type and amount of temporary leak sealant that could be used. If
additional sealant was necessary, an engineering evaluations was
typically required. TMM-031, Evaluation of On-Line Flowable Packing,
provided additional guidance for evaluating the temporary leak sealant.
As required by the temporary modification process, all temporary
modifications were subject to a 10 CFR 50.59 safety evaluation.
Furthermore, temporary modifications to safety related equipment were
reviewed as design changes. Typically QC was involved in the review
process when a "Q" component was affected. Temporary modifications to
non-safety related equipment received a technical review. Each
temporary modification was reviewed and approved by senior plant
management before being implemented. As required by the temporary
modification process, temporary leak sealants were required to be
replaced at the next refueling outage.
Battery Post Fractures
While performing routine maintenance of the "B" battery during RF015,
the licensee detected that the battery posts on many of the cells were
deformed, and some of the terminal posts had experienced fractures.
One of the cells of the "B" battery (cell 31) was sent to the Harris E&E
Center, where the posts were examined. Some subsurface fracturing was
observed, with evidence of oxidation along the fracture lines.
This condition raised concerns over the acceptability of the battery
with respect to meeting the electrical requirements of providing the
required output for safety-related functions, and the seismic integrity
of the battery. The "B" battery is a safety-related battery, which
supplies power to the "B" train of safety-related equipment, as well as
some non-safety related loads. The "B" safety train is one of the two
redundant trains of equipment designed to safely shut the plant down in
case of an accident. Cell 31 is one of the original cells installed in
the battery in 1978 and was chosen as representing the most significant
fracturing visually observed on the battery. The positive post of Cell
31 was determined to have had random fractures extending over
approximately 90 percent of a cross-sectional area for a section taken
through the bolt hole parallel to the top surface of the battery cell.
The cause of the fracturing is apparently due to over-torquing.
The major concern is that the fractures and deformation of the posts
could cause an increase in the resistance of the intercell connections.
This would result in a reduction in the load carrying capability of the
battery and a reduction in voltage from individual cells, which would
result in an overall reduction in the battery voltage.
13
The capability of the battery to support the electrical load
requirements is documented during refueling outages by the performance
of either MST-920 or MST-921. MST-921 is a service test which tests the
load profile for the battery for a one hour duration. This test is done
every refueling outage. MST-920 is a performance test, which tests the
performance of the battery with a continuous load for an eight hour
duration. MST-920 is performed every five years.
MST-920 was performed on the "B" battery on September 26, 1993, and
showed that the "B" battery had a capacity of 103.1 percent. This test
was performed before the fracturing was observed. The licensee stated
that this test demonstrated that the battery capacity had not been
adversely impacted by the fracturing. MST-920 was also performed during
RF013 in 1988 and showed a capacity of 102.1 percent.
Preventative Maintenance Procedure, PM-411 measures the resistance of
the intercell connections with an acceptance level of less than 50
micro-ohms across the strap connections. These tests are consistently
below the acceptance criteria and have not shown an increasing trend
toward higher resistance levels. PM-411 also covers torquing values for
the intercell connections. Torque values for these connections did not
exist prior to 1987.
The licensee stated that the effect of battery cell-to-cell seismic load
transfer is not addressed by design basis documents for the Robinson
site. However, considerable information is available to demonstrate
that there is little likelihood of battery cell failure due to battery
cell terminal post or connecting strap malfunction. The licensee stated
that this conclusion is valid even considering the as-found condition of
the "B" battery cells at Robinson.
The licensee concluded that the battery will perform its safety
functions with the observed fractures in the posts.
This issue is being referred to the Region II Division of Reactor Safety
for followup and will be tracked as IFI 93-21-06: Vital Battery Terminal
Fractures.
For the maintenance observations no violations or deviations were
identified. One inspector followup item concerning battery post
fractures was identified.
5. Followup (92700, 92701, 92702)
(Closed) URI 93-19-05, Ventilation System Damper Manipulation During
Performance Of Surveillance Testing. Inspection Report 93-19 discusses
manipulation of air cleaning unit dampers on May 1, 1992, during
Engineering Surveillance Test Procedure, EST-023, Control Room Emergency
Ventilation System. As documented in the completed EST, this
manipulation was required to eliminate backflow through the idle ACU
fans due to the backdraft dampers not fully shutting. When questioned
on the appropriateness of this action, the system engineer stated that
14
counterweights had been added to the dampers to eliminate this problem.
However, the system engineer was unsure if these additional weights were
added before or after the May 1, 1992, performance of EST-023.
On September 27, 1993, the system engineer informed the inspectors that
the damper counterweights had been added approximately 5 months prior to
the May 1 performance of the EST. Hence, the counterweights failed to
ensure that the backdraft dampers shut during performance of the EST.
The system engineer stated that no further corrective actions, beyond
manipulating the dampers during the test had been performed.
The licensee also stated that this manipulation would not be necessary
with an ACU fan operating since the fans aid in damper closing. An ACU
fan starts when the ventilation system shifts to its safeguards
configuration.
Additionally, the system engineer pointed out that the operation of
these dampers is also checked every two weeks during the performance of
Operations Surveillance Testing, OST-750, Control Room Emergency
Ventilation System Test and that failure of these dampers to operate has
not been observed. Based on this information and observations by the
inspectors of control room ventilation system operation, the inspectors
concluded that there was no safety significance to the initial damper
manipulation.
(Closed) IFI 91-16-01, Review Methods to Achieve Criticality in the
Source Range. The SRNIs and IRNIs were retracted about two feet as part
of corrective actions for ACR 91-285. A decrease in SRNI indication by
about a factor of two was observed. The inspector reviewed EST-050,
Refueling Startup Procedure, completed for Cycle 14 and Cycle 15
startups.
The inspector compared the data taken during initial
criticality by boron dilution and found that criticality for Cycle 15
was achieved about a decade below the P-6 IRNI Low Flux Reactor Trip
block and SRNI de-energization setpoint. However, because new IRNI
detectors were installed during RFO 14, no meaningful comparison between
Cycle 14 and Cycle 15 data could be made. The inspector noted that the
POAH value for Cycle 15 was about a full decade below the value for
Cycle 14.
The inspector concluded that retracting the SRNIs and IRNIs
would allow criticality to be achieved before de-energizing the SRNIs.
(Closed) IFI 91-16-02, Review Enhancements to Procedure EST-050 to
Increase Guidance. This IFI dealt with additional guidance in EST-050
for activities associated with determining the POAH and establishing a
margin for the ZPTR, MTC calculation, and reactivity computer
calibration accuracy. The inspector reviewed EST-050, Revision 12, for
changes in response to this IFI.
A note was added before step 7.1.20 which listed plant parameter changes
that would indicate the POAH. These parameters included an increase in
pressurizer level, RCS T.,,
or RCS TH., or a decrease in reactivity
computer flux or reactivity. The inspector discussed these indications
15
with the licensee to verify which indication would be the most
responsive. The licensee stated the reactivity computer would respond
first due to the increased sensitivity to small flux changes. The
inspector then questioned if the other indications were sufficiently
responsive. The licensee stated that pressurizer level was the least
responsive indicator, but RCS T.,9 and RCS THt responded quickly because
the bypass manifold had been removed. Based on this review, the
inspector concluded these parameter changes were acceptable as
indications of the POAH.
EST-050 had been revised to establish the ZPTR Upper Limit up to one
decade below the POAH. The inspector reviewed the ZPTR limits
determined for the Cycle 15 startup and discussed these limits with the
licensee. The licensee stated that a 3/4 ZPTR decade Upper Limit was
used due to insufficient S/N for a full decade. The reduced S/N was
evidently due to lower neutron leakage for this particular core loading.
The inspector concluded that the one decade established sufficient
margin between the POAH and the ZPTR Upper Limit.
To provide additional resolution for the reactivity-temperature trace
used for MTC determination, an X-Y plotter was used to record the
reactivity-temperature data. Also, EST-050 directed the plotter scaled
to the maximum spans available for the paper size being used. The
inspector concluded the use of an X-Y plotter and maximizing the span
would enhance the ability to analyze the reactivity-temperature data.
The amount of data to be taken was specified as up to a 5*F change or
three or more inches of trace length. Because a heatup or cooldown rate
of 10*F/hr was required, the inspector questioned if trace length was
sufficient to obtain an acceptable reactivity-temperature relationship.
The licensee stated that the three inches of trace length historically
represented about a 4F temperature change. The inspector concluded the
specified trace length was an acceptable indication that sufficient data
was obtained.
Also EST-050 Attachment 8.4 was changed to require the two reactivity
computer calibration data set averages agree within +/-4%. The inspector
concluded this requirement was sufficiently accurate.
(Closed) IFI 93-12-05, NRC Review and Follow-up of Any SWS Heat
Exchanger Inspections and Tests During the 1993 Refueling Outage. The
inspector reviewed photographs of the EDG 'A' heat exchangers, the CCW
'A' heat exchanger, and the AFW 'A' lube oil cooler. These photographs
were taken immediately after these components were opened to record the
as-found condition.
The end bells for the EDG 'A' heat exchangers had soft sludge deposited
in defined rows. Also, nodules of the same soft sludge had deposited in
the water box of the heat exchangers. This sludge was easily removed and
there was insufficient accumulation to interfere with water flow through
the heat exchangers. There was a noticeable decrease in the amount of
sludge deposited when compared to the photographs for the previous
16
inspection. Small pebbles of manganese dioxide were found inside
several tubes. These pebbles did not obstruct flow. Some scale had
developed inside the tubes and some tubes were plugged with soft sludge.
The scale and soft sludge were removed during mechanical cleaning.
Prior to the heat exchanger final flush and closure, the inspector
checked the condition of the tubes.
Based on the visual examination,
the tubes were clear of any plugging, but some pebbles of manganese
dioxide were present. The licensee said these would be removed by the
final flush.
The AFW lube oil coolers were a four-pass heat exchanger in a four foot
long by six inch diameter jacket. The lube oil coolers were found to
have substantial fouling. The end bell was 90 percent fouled as were 30
percent of the tubes. The AFW 'A' lube oil cooler failed hydrostatic
testing and was replaced. The AFW 'B' lube oil cooler was mechanically
cleaned and reinstalled. The inspector reviewed the data for the last
six performances of OST-201, Motor Driven Auxiliary Feedwater System
Component Test, for indications of lube oil cooler degradation. The
lube oil differential temperature, which remained between 5'F and 7*F
over the last six performances, did not indicate any degradation. The
inspector discussed the substantial fouling of the lube oil coolers with
the licensee. The licensee stated the lube oil coolers historically
have had substantial fouling with no degradation indicated by OST-201.
The photographs of the CCW 'A' heat exchanger indicated minimal fouling.
Small quantities of the soft sludge had accumulated only in pockets and
crevices in the coating of the water box. The tubes appeared to be
clear of any fouling. Subsequent inside dimensional testing was
inconclusive and will be the subject of future inspection effort.
(Closed) LER 93-10, Diesel Generator Fire. At approximately 3:00 p.m.
on August 16, 1993, a small oil fire occurred on the exhaust manifold of
"A" Emergency Diesel Generator (EDG) during the performance of OST-401,
Emergency Diesels Slow Speed Start. The Unit was operating at 100
percent power at the time of the event. The fire was immediately
extinguished by the operator using a portable fire extinguisher. The
"A" EDG continued to operate for the period of time required by the OST
and the plant continued to operate at 100 percent power.
At approximately 3:20 p.m. the licensee declared an Alert, based on a
fire with the potential to affect safety-related equipment. The
Emergency Response Organization (ERO) was notified and the Technical
Support Center (TSC) was activated. Appropriate notifications were made
to State and Counties, the NRC, and other organizations as required.
Based on the fire being extinguished and the "A" EDG continuing to
operate in a loaded condition, the Alert was downgraded and the
emergency terminated at 4:37 p.m..
The licensee formed an incident evaluation team to determine the cause
of the fire and to recommend corrective actions. The incident
17
evaluation team identified that the gasket installed on the exhaust
manifold had been misaligned and crimped allowing oil from the pre-lube
process to leak into the heat shield where it later ignited and caused a
small fire.
The exhaust manifold gasket was replaced and the "A" EDG tested
satisfactorily. The licensee has committed to revising procedure CM
640, EDG Exhaust System Maintenance, to include more specific guidance
on the installation of the exhaust manifold gaskets. The procedure
change is to be completed by November 12, 1993.
6. Exit Interview (71701)
The inspection scope and findings were summarized on October 15, with
those persons indicated in paragraph 1. The inspectors described the
areas inspected and discussed in detail the inspection findings listed
below and in the summary. Dissenting comments were not received from
the licensee. The licensee did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during this
inspection.
Item Number
Description/Reference Paragraph
93-21-01
NCV: Failure Of STA To Conduct Turnover Required
By OMM-008 (Paragraph 3)
93-21-02
VIO: Failure To Follow Procedures, Two Examples
Concerning Opening The Incorrect Electrical
Breaker, and Sludge Lance Rig Vent Valve
(Paragraph 3)
93-21-03
NCV: Area Fire Watch Vacates Post (Paragraph 3)
93-21-04
VIO: Failure To Maintain Containment Integrity
During Refueling Operations (paragraph 3)
93-21-05
IFI: Fuel Assembly Loose Part (Paragraph 3)
93-21-06
IFI: Vital Battery Terminal Fractures (Paragraph
4)
7.
List of Acronyms and Initialisms
ACR
Adverse Condition Report
ACU
Air Cooling Unit
Component Cooling Water
CFR
Code of Federal Regulations
Corrective Maintenance
CV
Containment Vessel
EE
Engineering Evaluation
18
Environmental Qualification
Emergency Response Organization
EST
Engineering Surveillance Test
Fire Protection
General Procedure
IFI
Inspector Followup Item
IR
Inspection Report
IRNIs
Intermediate Range Nuclear Instruments
LCO
Limiting Condition for Operation
LER
Licensee Event Report
Loss of Coolant Accident
Motor Control Center
Maintenance Surveillance Test
MTC
Moderator Temperature Coefficient
CV
Non-cited Violation
OMM
Operations Management Manual
ONCA
Off Normal Condition Analysis Form
OP
Operations Procedure
OST
Operations Surveillance Test
Preventive Maintenance
Point Of Adding Heat
RCCA
Rod Control Cluster Assembly
REV
Revision
Refueling Outage
Refueling Water Storage Tank
S/G
Safety Injection
S/N
Signal-to-Noise Ratio
Special Procedure
SRNIs
Source Range Nuclear Instruments
Service Water System
TAVG
Temperature Average
TMM
Technical Support Management Manual
TS
Technical Specification
WR/JO
Work Request/Job Order
ZPTR
Zero Power Testing Range