ML14181A694
| ML14181A694 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/19/1995 |
| From: | Ogle C, William Orders, Verrelli D NRC Office of Inspection & Enforcement (IE Region II) |
| To: | Carolina Power & Light Co |
| Shared Package | |
| ML14181A692 | List: |
| References | |
| 50-261-95-12, NUDOCS 9505300110 | |
| Download: ML14181A694 (20) | |
See also: IR 05000261/1995012
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION 11
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report No.:
50-261/95-12
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: .March 19 - April 22, 1995
Lead Inspector:
<2'>c
7,
/
W. .
rderf, Seni~r Re ident
Date Signed
Other Inspector: c
'
) 7, /(
1/
C.
Ogle, Res'd nt I"pector
Date Signed
Approved
by:__
/___
David/M.
Verre li, Chief
Date Signed
Reactor Proje s 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 March 20, 21, 25, 26, 27, 30, and April 2, 3, 4, 5, 6,
9, 10,
12,
17,
18, and 20,
1995.
RESULTS:
Plant Operations:
One Violation with two examples was identified. The first example involved the
licensee's failure to verify the configuration of the Residual Heat Removal
System as required by procedure OST-254. The second example involved a
manager's failure to brief the on-shift operating crew prior to the
performance of OST-254.
Maintenance:
.
One Violation was identified concerning the licensee's failure to specify
adequate post maintenance testing.
9505300110 950519
PDR ADOCK 05000261
Q
2
A Non-Cited Violation was identified involving the licensee's failure to
include pin feeler gauges and shop-fabricated feeler gauges in the Measuring
and Test Equipment Program.
Engineering:
One Non-Cited Violation was identified involving the alteration of test data.
Plant Support:
One Violation was identified concerning the licensee's failure to maintain
control of Safeguards Information.
REPORT DETAILS
PERSONS CONTACTED
Licensee Employees:
- J. Boska, Manager, Electrical/Instrumentation and Control
W. Brand, Supervisor, Environmental Radiation Control
M. Brown, Manager, Design Engineering
A. Carley, Manager, Site Communications
G. Castleberry, Manager, Plant Electrical Engineering
- B. Clark, Manager, Maintenance
- D. Crook, Senior Specialist, Licensing/Regulatory Compliance
- R. Dayton, Outage Management
C. Gray, Manager, Materials and Contract Services
D. Gudger, Senior Specialist, Licensing/Regulatory Programs
- W. Hatcher, Director, Nuclear Security
- C. Hinnant, Vice President, Robinson Nuclear Project
P. Jenny, Manager, Emergency Preparedness
- K. Jury, Manager, Licensing/Regulatory Programs
J. Kozyra, Licensing/Regulatory Programs
- R. Krich, Manager, Regulatory Affairs
- F. Lowery, Manager, Work Control
- E. Martin, Manager Document Services
- B. Meyer, Manager, Operations
- G. Miller, Manager, Robinson Engineering Support Section
- J. Moyer, Manager, Nuclear Assessment Section
- W. Randlett, Manager, Security
- 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
- Attended exit interview
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2
2. PLANT STATUS AND ACTIVITIES
a.
Operating Status
The unit operated at or near full power for the entire report
period.
On April 11, 1995, the licensee declared an Unusual Event
when RCS unidentified leakage exceeded 10 gallons per minute
during the performance of an operations surveillance test.
Details of this event are delineated in paragraph 3.
b.
Other NRC Inspections and Meetings
Three Region II based inspections were conducted during the report
period. The first inspection, conducted during the week of
March 27-31, 1995, was performed by M. Ernstes and C. Payne.
Their inspections results are documented in 50-261/95-10. A
second inspection was performed by W. Loo during April 3-4, 1995.
His inspection results are documented in report 72-3/95-01.
The
third inspection, conducted during the week of April 3-7, 1995,
was performed by J. Kreh. His inspection results are documented
in report 50-261/95-11.
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 status,
and review of facility records.
The inspectors reviewed shift logs, operation's records, data
sheets, instrument traces, and the equipment malfunctions record
to assess equipment operability and compliance with TS. The
inspectors evaluated the operating staff to determine if they were
knowledgeable of plant conditions, responded properly to alarms,
adhered to procedures and applicable administrative controls, and
were cognizant of in-progress surveillance and maintenance
activities. The inspectors performed instrument channel checks,
reviewed component status, and assessed safety-related parameters
to determine conformance with TS.
Shift changes were routinely
observed to determine that system status continuity was maintained
and that proper control room staffing existed. Access to the
control room was controlled in the main, and operations personnel
carried out their assigned duties in an effective manner. Control
room demeanor and communications were appropriate.
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,
3
physical protection controls, and equipment tagging procedures
were properly implemented.
b.
Onsite Response to Events (93702)
Event Summary
At approximately 10:15 p.m. on the evening of April 11, 1995, the
licensee attempted to perform OST-254, Residual Heat Removal
System and RHR Loop Sampling System Leak Test. The surveillance
test uses the letdown portion of the Chemical and Volume Control
System to pressurize the RHR system which is then checked for
leakage. After pressurizing the "A" train of the RHR system,
operators detected an unidentified reactor coolant leak of 24.8
gallons per minute. The operators terminated the test and
attempted to isolate the leak by closing valve HCV-142, the valve
which had been opened to pressurize the system. The leakage was
reduced to 15.6 gallons per minute but was not stopped.
Ultimately, the leakage was terminated by closing manual isolation
valve, CVC-205B, which is in series with HCV-142. The licensee
declared an Unusual Event at 11:00 p.m. due to -eactor coolant
system leakage greater than 10 gallons per minute. The event was
terminated at 11:33 p.m. after confirmatory tests indicated that
RCS leakage was within specifications.
Background
OST-254, Residual Heat Removal System and RHR Loop Sampling System
Leak Test, was an annual surveillance test performed to meet the
requirements of Technical Specification 4.4.3. The purpose of the
test was to determine external system leakage of those portions of
the RHR System outside containment. The test had last been
performed on January 17, 1994, and had to be performed by
April 19, 1995, to fulfill the surveillance requirement. The
procedure contained methods for performing the test while the unit
was either at power or shutdown.
On April 11, 1995, the unit was operating at 100 percent power.
Accordingly, the licensee employed Section 7.1 of the procedure,
RHR in Standby for Low Pressure Safety Injection, in their attempt
to perform the test. This section tests one train of RHR at a
time by pressurizing that train and measuring system leakage while
maintaining the alternate train of RHR operable and aligned for
low pressure injection. After the system is aligned, the train to
be tested is pressurized to 350-370 psig from the CVCS by opening
valve HCV-142. Once the system has been pressurized, valve SI
861A, CV Sump to RHR, is opened to pressurize the space between it
and valve SI-860A. This action creates a known RCS leak to the
containment sump due to a 3/16 inch hole which has been drilled in
the upstream disc of valve SI-860A. SI-861A is closed once the
required leakage measurements are completed on and around the two
valves. After re-closure of SI-861A, the remainder of that
4
train's components are checked for leakage. When all leakage
measurements have been completed, the system is realigned, the RHR
pump is run for operability verification, and train "B" is then
tested using the same methodology.
Event Details
The licensee determined that performing the test at power was a
Case Two activity pursuant to procedure PLP-037, Conduct of
Infrequently Performed Tests or Evolutions. As such, the
personnel who were to perform the evolution were to receive a pre
job briefing from a Management Designated Monitor. This briefing
was completed at 7:30 p.m. on April 11, 1995. The Management
Designated Monitor briefed the extra personnel who had been
assigned to perform the test, but did not brief the operations
crew on shift.
Step 5.3.2.2 of PLP-037 requires that "The MDM
shall brief the operating and test personnel...prior to the
performance of the test...."
The Management Designated Monitor's
failure to brief the operating crew is a Violation and is one of
two examples which collectively constitutes VIO 95-12-01,
Operations Failure to Follow Procedure During OST-254.
Topics discussed during the briefing included a number of related
industry events but did not include a previous Robinson event
which occurred on March 19,.1994, in which an RCS leak was
initiated due to failure of HCV-142 to isolate flow. Furthermore,
the Management Designated Monitor did not delineate in writing the
duties of the additional operating crew assigned to OST-254 as
suggested in PLP-037. The Management Designated Monitor's failure
to include the March 19, 1994, event in the briefing and his
failure to delineate, in writing, the duties of the additional
operating crew are considered deficiencies in his implementation
of the procedure.
At 9:30 p.m., on the evening of April 11, 1995, the first active
portion of OST-254 began with running the RHR pumps to remove any
trapped air in the lines. At 10:15 p.m., HCV-142 was throttled
opened to pressurize the A train. Based on ERFIS data, VCT level
began to decrease almost immediately. The operators felt that
this indication was normal and reflected the inventory required to
pressurize the system. A review of the ERFIS data indicated that
RWST level began increasing almost immediately also but since the
operators were not monitoring the RWST on ERFIS, they were unaware
of this trend. OST-254 CAUTIONS the operator in step 7.1 that
RWST level should be monitored for an unexplained increase in
level, but does not provide guidance relative to how to
effectively monitor for level changes. As a result, the operators
monitored RWST level on a control room gauge which did not have
the accuracy necessary for effectively detecting small level
changes.
This is considered a procedural deficiency.
5
By 10:30 p.m., HCV-142 was full open and RHR pump "A" discharge
pressure was verified to be in the control band of 350-370 psig.
At 10:42 p.m., valve SI-861A, CV Sump to RHR, was opened.
Operators knew that opening this valve would create a leak,
because valve SI-860A, which is located immediately downstream of
valve SI-861A, has a hole drilled in its upstream disc. The
magnitude of this leak rate had not been calculated. The fact that
the quantity of this known leak had not been calculated and
included in the procedure is considered a procedural deficiency.
At 10:33 p.m. a manual makeup to the VCT of 150 gallons was
required. The operators believed this, also, to be an expected
response, as the RHR system pressurized/stabilized.
From 10:32 p.m. to 10:38 p.m., the area between and around valves
SI-860A and SI-861A was inspected for external leakage. At
10:38 p.m., SI-861A was closed. After this valve was closed,
operators observed that the VCT level was still trending down.
According to the licensee, it was at this time that the operators
realized that a leak existed. The operators spent the next ten
minutes diagnosing the event as well as performing a 5 minute leak
rate calculation. At 10:48 p.m., the leak rate was determined to
be 24.8 gpm. AOP-016, Excessive Primary Plant Leakage, and
Technical Specification Action Statement 3.1.5.1 on RCS leakage
were entered.
Between 10:40 p.m. and 10:53 p.m., three automatic makeups to the
VCT occurred, each of 100 gallons.
At 10:53 p.m., valve HCV-142 was closed. Since the valve leaked
by, the leak was not terminated but was decreased to 15.6 gpm. At
10:55 p.m. valve CVC-205B which is upstream of HCV-142, was closed
which isolated CVCS from RHR, and terminated the event.
At 10:56 p.m., VCT level stabilized which indicated that leakage
had returned to normal, and Technical Specification Action
Statement 3.1.5.1 on RCS leakage was exited.
At 11:00 p.m., the licensee declared an Unusual Event based on
unidentified leakage being in excess of 10 gpm.
At 11:10 p.m., 300 gallons of primary water were added to the VCT.
This addition brought the total to 750 gallons of water which was
required to bring the VCT level back to pre-event status.
At 11:13 p.m., the licensee exited AOP-016, Excessive Primary
Plant Leakage, and at 11:33 p.m., the licensee terminated the
Unusual Event.
6
Licensee Corrective Actions
The licensee formed an Event Review Team, and a procedure revision
team the next morning. The Event Review Team did a thorough job
in evaluating the event. The Event Review Team concluded that:
approximately 673 gallons of reactor coolant inventory was
transferred to the RWST
the leak rate through SI-860A was approximately 17 gallons
per minute
the emergency core cooling system remained operable
throughout the event
there was no radiological release
OST-254 did not contain specific procedural guidance to
address a potential RCS leak
the expected system response to the initial opening of
HCV-142 was not quantified
RWST level indicator band is too wide to detect a level
change of several hundred gallons
no estimate of the known leak rate through SI-861A was
provided
HCV-142 history was not factored into the procedure
the performance of PLP-037 was deficient
the failure to include the normal on-shift crew in the
Case 2 briefing did not meet the requirement of PLP-037
the Robinson event of March 19, 1994, was not discussed
the duties, authority and responsibility of "the extra
personnel" were not specified in writing, as required by
PLP-037
the operation of HCV-142 has been erratic, and previous
corrective actions to repair it have not been effective
performance of OST-254 at power is not recommended
the Operating crew showed conservative decision making in
responding to this leak
The procedure revision team was tasked to evaluate the event, and
determine if OST-254 could be performed in a more conservative
7
manner at power. In parallel with these efforts, the licensee
also submitted an Emergency Technical Specification change request
to allow the surveillance to be performed on a refueling basis as
opposed to an annual basis, as was the case. Ultimately, the
technical specification change was granted, and the procedure
revision team was disbanded.
The licensee evaluated the RHR system configuration to determine
the possible leak paths from the system to the RWST. This review
identified eight possible combinations of component leakage paths
which could have resulted in the observed result. One of these
possibilities was valve RHR-7570, the B RHR heat exchange bypass
valve, which was found to be 1/4 turn open.
OST-254 requires in step 7.1.2 that operators "Verify the RHR
System is aligned for standby low pressure injection IAW (in
accordance with) OP-201.
OP-201, Residual Heat Removal System
requires in step 6.2.2.1 that when the system is aligned for low
pressure injection, valve RHR-757D is to be locked closed.
On April 11, 1995, valve RHR-757D was found 1/4 turn open. This
is the second of two examples which collectively constitutes
VIO 95-12-01: Operations Failure to Follow Procedure During
OST-254.
Conclusions
The event did not result in an offsite radiological release, did
not render the emergency core cooling system inoperable, and was
dealt with in an adequate manner.
RESULTS:
One Violation with two examples was identified in this evaluation area.
The first example involved a manager's failure to brief the on-shift
operating crew prior to the performance of OST-254. The second example
involved the licensee's failure to verify the configuration of the
Residual Heat Removal System as required by procedure OST-254.
The operating crew on shift during the attempted performance of OST-254
responded adequately to the event.
The leak was effectively isolated
and the appropriate contingency procedures employed.
Procedure adequacy and procedure adherence continue to be a persistent
problem. Other than the specified incidents referenced above, the
operations program was effectively implemented.
8
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:
CM-610
Install New Rings In Number 10
Cylinder For EDG A
WR/JO 95-AHMCOO1
Inspection and Testing Of Circuit
Breakers For 480V Bus E-2 (SW Pump C
Only)
WR/JO 95-ADKQ1
Replace Lights In Service Water Pump
C Local Controller
Non-Controlled Measuring Equipment Used In Breaker Maintenance
The inspectors witnessed a portion of the preventive maintenance
performed on the SW pump C breaker on April 2, 1995. During the
post-maintenance review, the inspectors questioned the licensee's
control of pin gauges which were present at the job site during
the maintenance. As described by maintenance personnel and as
discussed in Corrective Maintenance Procedure, CM-305,
Westinghouse "DB" Type Circuit Breakers Maintenance, these gauges
are used to verify proper tolerances on various breaker component
clearances. The inspectors noted that the pin gauges were stored
in vials segregated by pin diameter. The vials were marked as to
pin diameters, but neither the pins nor the vials bore a M&TE
identification sticker. Only two of the twelve pins were
individually marked with their diameter. The inspectors were
concerned that given this limited control, nothing prevented a pin
of incorrect diameter being used, given an inadvertent storage of
a pin in the wrong vial.
The inspectors subsequently measured
several of the pins using a micrometer and verified that at the
time of the measurement, that no pins were stored in the wrong
vials.
In response to these observations, the inspectors reviewed the
licensee's Plant Program Procedure, PLP-053, Measuring and Test
Equipment (M&TE), the Corporate QA Manual, as well as American
National Standard ANS-3.2, Administrative Controls and Quality
Assurance for the Operational Phase of Nuclear Power Plants.
9
On April 6, 1995, during a follow-up field observation, the
inspectors observed four shop-fabricated feeler gauges at the job
site for maintenance on the HVH-2 breaker. The inspectors were
advised that these gauges were routinely used to verify the
"GAP G" on this style breaker. These gauges were fabricated with
long shanks to prevent personnel injury which could result if the
breaker actuated while this gap was being measured.
(The pin
gauges are much shorter and their use requires the technicians'
hand be nearer the breaker.)
The inspectors noted that 3 of the 4
shop-fabricated feeler gauges were not marked as to their size and
none were labelled as controlled M&TE.
Following an inspector
request, the feeler gauges were measured.
(The inspectors were
subsequently informed that the Maintenance Manager had also
directed a similar measurement be taken.)
The inspectors were
advised that the following results were obtained for the four
feeler gauges:
0.094 inch nominal size
Tool 1 = 0.095 inch
Tool 2 = 0.092 inch
0.050 inch nominal size
Tool 3 = 0.048 inch
Tool 4 = 0.051 inch
These measurements were taken with a micrometer with a reported
accuracy of + 0.001 inch.
Based on these results, the inspectors concluded that, as
measured, two of the four tools were not accurate enough to
satisfy the gap measuring requirements specified in CM-305.
Furthermore, even if the uncertainty in the M&TE device used to
measure the field fabricated feeler gauges was applied non
conservatively, one gauge would still not be accurate enough to
ensure that the gap measured on the breaker was within the
tolerances specified in the procedure. The potential non
conservatism introduced by the feeler gauge inaccuracy would be
small.
However, the inspectors noted that had the field
fabricated feeler gauges been included in the M&TE program, this
tool inaccuracy may have been detected earlier.
Following these observations, the pin feeler gauges were measured
by the licensee and verified to be stored in the proper vials.
The licensee also stated their intention to place the pin feeler
gauges in the calibration program. Additionally, the shop
fabricated feeler gauges have been removed from service and
replaced with tools for which the calibration can more readily be
verified. Overall, the inspectors concluded that the failure to
control the pin feeler gauges and the shop-fabricated feeler
gauges was contrary to the requirements of Plant Program
Procedure, PLP-053, Measuring and Test Equipment (M&TE)
Calibration Program. However, this NRC 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
10
Non-Cited Violation, NCV 95-12-02, Breaker Clearance Measuring
Tools Not Properly Controlled.
HVH-2 Breaker Retest
On the morning of April 6, 1995, HVH-2 was removed from service,
in part, to accomplish breaker PMs.
Material deficiencies were
discovered with the HVH-2 breaker during the PMs and a replacement
breaker was installed. At approximately 7:40 p.m., that day,
while observing the restoration of HVH-2 to service, the
inspectors questioned the adequacy of the specified post
maintenance test. The inspectors were concerned that the PMTR
accomplished by Operations, namely shutting the breaker from the
RTGB, failed to adequately verify the operability of the
replacement breaker. Specifically, the inspectors questioned the
need to verify that the breaker would also open and that closing
would occur within a timeframe which would not impact the sequence
of other loads onto the E-bus during an accident. These concerns
were based on the inspectors' review of Maintenance Management
Manual Procedure, MMM-003, Appendix A, Post Maintenance Testing.
This procedure specifies that post-maintenance testing demonstrate
that the breaker responds properly to all demand signals.
Further, the procedure specifies that breakers on the E-1 and E-2
busses that get an autostart signal from a safeguards actuation,
have a breaker time test performed as PMTR.
In response to these concerns, the HVH-2 breaker was opened from
the RTGB. Additionally, the licensee timed the operation of the
replacement breaker from the RTGB. In accordance with a
subsequent ESR, this measured breaker time was satisfactory.
A
CR was also generated regarding this issue.
In response, the inspectors reviewed the WR/JOs used to accomplish
the breaker repairs; MMM-003, Appendix A; and the CR. The
inspectors also reviewed the ESR produced to investigate the
breaker closing timing requirement for the E-buss breakers.
Additionally, the inspectors interviewed the planner who
designated the original PMTR. The inspectors witnessed the timing
test performed to demonstrate that the breaker was operable.
Based on this review, the inspectors determined that the planner
failed to properly establish the PMTR as specified by MMM-003,
Appendix A. This planner advised the inspectors that he failed to
consult MMM-003, while planning the work.
Instead he relied on
similar historical PMs and existing, related procedures for
breaker maintenance. This is contrary to the requirements of
MMM-003, Appendix A. This is identified as Violation VIO 95-12
03, Maintenance Planner Fails to Properly Develop Breaker PMTR.
Subsequent to the inspectors observation, the licensee developed
an ESR to evaluate the need to perform a timing test following
breaker replacement. This ESR concluded, that the acceptance
timing test can be as simple as personnel observation that the
breaker closed immediately or verified to be less than 1/2 second.
While this conclusion diminished the safety significance of the
unperformed timing test, it does not negate the fact that the
planner failed to follow process by not consulting MMM-003,
Appendix A.
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
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 activity:
EDG Run In After Maintenance
c.
Followup - Maintenance (92902)
(Closed) URI 93-19-03, Adequacy of Testing for CCW Pump Autostart
Feature
URI 93-19-03 documented the inspectors' concerns associated with
the calibration procedure associated with the CCW pump discharge
pressure switch. The inspectors were concerned that the observed
testing failed to adequately test the circuitry involved.
The inspectors reviewed a Nuclear Engineering Department memo
dated January 7, 1994, which addressed these concerns. This memo
provides justification for the licensee's testing strategy. The
inspectors reviewed the memo and have no further question on this
issue. Therefore, URI 93-19-03 is closed.
RESULTS:
One Violation and one Non-Cited Violation were identified in this
evaluation area.
The Violation concerns the licensee's failure to specify adequate post
maintenance testing. The Non-Cited Violation involved the licensee's
12
failure to include pin feeler gauges and shop-fabricated feeler gauges
in the Measuring and Test Equipment Program.
The inspectors observed selected safety-related maintenance activities
on diverse systems and components. With the exception of the events
discussed above, these activities were conducted in accordance with TS,
approved procedures, and appropriate industry codes and standards. The
maintenance program was effectively implemented.
5.
ENGINEERING (37551)
The inspectors evaluated selected engineering related events to
determine the effectiveness of the licensee's controls in identifying,
resolving and preventing issues by reviewing areas such as the
corrective action system, root cause analysis, safety committees, and
self assessment.
(Closed) URI 93-28-03, Alteration of OST-410 Data
URI 93-28-03, Alteration Of OST-410 Data, concerned an event involving
the alteration of test data during the performance of OST-410, Emergency
Diesel Generator "A" Twenty-Four Hour Load Test on October 31, 1993.
The responsible individual acknowledged changing the 2510 KW reading to
2500 KW, after operations voiced concern over the reading. The
individual based his actions on his belief that the EDG A KW meter could
not be read to an accuracy of 10 KW.
The inspectors reviewed the altered data sheet and noted that no
initials or other discriminating marks were provided to indicate that
the data had been altered. Subsequently, the inspectors reviewed an
evaluation performed by Engineering Technical Support which concluded
that there was no technical significance to this brief excursion.
Records management procedure RMP-001, Records And QA Records Storage,
requires in section 4.1.4. that corrections to QA records, such as
completed procedures, be made by drawing a single line through the
incorrect information, writing the correct information adjacent to the
deletion and initialing and dating the correction. The responsible
individual failed to follow this procedure requirement. 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 95-12-04, Alteration of Test Data.
Unresolved Item 93-28-03 is closed.
(Closed) URI 94-27-09, EOF/TSC Ventilation System
URI 94-27-09 deals with potential deficiencies in the EOF/TSC
ventilation system observed by the inspectors on November 15, 1994.
These deficiencies included several uncapped cable penetrations in an
outer wall of the EOF/TSC mechanical equipment area; an out of service
13
iodine channel in the R-38 radiation monitor; and previous additions to
the EOF/TSC building which may have impacted the capability of the
building ventilation system to perform its function.
In response to the uncapped penetrations, the inspectors reviewed
Engineering Evaluation, EE 94-102, Evaluation of Penetrations in the EOF
Building. This evaluation was developed after similar concerns were
expressed by the inspectors in early 1994 following the initial
penetrations being made to the EOF/TSC outer wall.
This EE provides
information which demonstrates that the uncapped penetrations did not
violate the basis for the shielding analysis performed for the TSC/EOF.
The inspectors independently reviewed the EE and concluded that the EE
remained valid for the inspectors' observation on November 14, 1994.
Based on this, the inspectors concluded that the uncapped penetration
did not represent a meaningful degradation in the TSC/EOF.
The
inspectors have no further questions on this aspect of the URI.
The second concern identified in URI 94-27-09 involved the iodine
channel of the TSC/EOF radiation monitor (R-38) being carried in a
source check log as out of service. The inspectors were concerned that
this material deficiency could prevent automatic activation of the
EOF/TSC ventilation pressurization system. The inspectors reviewed
historical correspondence between the NRC and CP&L related to the
TSC/EOF. This correspondence specifically stated that the TSC
ventilation system need not automatically actuate as a result of alarm
conditions on installed monitoring equipment. The inspectors also noted
from a review of Plant Emergency Procedure, PEP-169, Radiological
Control Manager, that the EOF/TSC ventilation is placed on recirculation
whenever an Alert or higher is declared or the EOF is activated. Based
on this information, the inspectors concluded that the out of service
iodine channel had minimal safety significance. The inspectors have no
further questions on this aspect of the URI.
The third part of the URI dealt with the inspectors' questions regarding
potential modifications made to the EOF/TSC building and the impact
these additions had on the EOF/TSC ventilation system. In response to
these concerns the licensee generated several ESRs and conducted testing
of the EOF/TSC ventilation system. The inspectors reviewed these ESRs
and the results of the ventilation system testing. The inspectors also
interviewed the engineer involved in resolving this issue.
Additionally, the inspectors conducted an independent walkdown of the
EOF/TSC ventilation system. The inspectors also reviewed
correspondence, between the NRC and CP&L, related to the EOF/TSC
ventilation system.
Based on this review, the inspectors determined that the EOF/TSC was
constructed to maintain post-accident occupant dose below 5 Rem whole
body or its equivalent. The ventilation system incorporated a filter
system to remove contaminants from incoming air and was also designed to
maintain the building at a slightly positive pressure relative to
ambient conditions. The licensee's design was confirmed by an NRC order
dated February 21, 1984. Earlier NRC correspondence established the
14
guidelines for this design and specifically exempted portions of the
system from 10 CFR 40 Appendix B requirements.
Furthermore, the inspectors determined that two additions, identified as
Phase 1 and Phase 2 were made to the original EOF/TSC building.
Phase 1, added in September 1985, consisted of management and
administrative office space; Phase 2, added a simulator and training
offices to the building and was completed in December 1985. The
December 1985 additions were made outside the original EOF/TSC building
confines. Further, both of these additions included individual
ventilation systems which had no direct ties to the existing EOF/TSC
Ventilation system. Despite this independence, the Phase 2 addition
overlapped one of the original EOF/TSC entry vestibules.
(This entry
vestibule served as an air-lock boundary for the original EOF/TSC
building.)
The licensee was unable to provide any documentation that
acceptance testing had been performed following Phase 1 or Phase 2
construction to demonstrate that the original EOF/TSC ventilation system
performance, and hence, the capability of the system to maintain a
positive pressure, had been maintained. While an EST is routinely
performed to verify EOF/TSC ventilation system performance, this EST
does not consider the impact of the Phase 1 or Phase 2 ventilation
system.
On January 9, 1995, and again on January 17, 1995, the licensee
conducted testing of the EOF/TSC ventilation system. This testing
demonstrated that the original EOF/TSC would be maintained at a positive
pressure given that all fans in the Phase 1 and Phase 2 additions were
all on or all off. However, the EOF/TSC to outside differential
pressure did change as a result of changing the state of the fans in
these adjacent additions. The inspectors also consider it noteworthy
that with the EOF/TSC ventilation system HEPA filter partially
obstructed to near design differential pressure, and fans in Phase 1 and
Phase 2 in other than all on or off, the differential pressure between
the additions and EOF/TSC was effectively reduced to zero in some
situations. The licensee argued that this test was not an appropriate
test of the EOF/TSC ventilation. As understood by the inspectors, this
argument centered on the licensees contention that all fans on or off in
the additions was the design basis of the system. Further, the licensee
also contended that the near design differential pressure on the HEPA
filter represented an unrealistic test given the light dust loadings in
the area and the infrequent use of the system. Though these positions
are plausible, it was noteworthy that the licensee was unable to supply
historical design basis information to substantiate these positions.
While reviewing information related to this issue, the inspectors noted
that the ventilation model used in the EOF/TSC radiation shielding
analysis did not match the installed system configuration.
(The
inspectors were subsequently informed that a similar observation had
also been made by licensee personnel.)
The differences were primarily
related to the amount of filtered makeup air used and the incorporation
of a non-existent filtered recirculation flowpath into the model.
Following these observations, the EOF/TSC dose calculations were re-
15
performed by an off-site contractor. The inspectors reviewed the
results of these calculations and noted that the projected dose remained
below the original calculated values.
Overall, the inspectors concluded that the licensee retained the ability
to pressurize the EOF/TSC following the Phase 1 and Phase 2 additions
given the assumptions identified above. Hence, the licensee has
maintained the design of the EOF/TSC. However, based on the information
reviewed by the inspectors, it does not appear that maintaining this
design was the result of a controlled process.
Further, shortcomings
were identified in the licensees understanding of the design basis of
the system. These deficiencies are identified as a weakness. The
inspectors have no further questions on this aspect of the URI.
URI 94-27-09 is closed.
(Closed) URI 95-06-04, Potential Design Vulnerability of Selected
Safety-Related Circuits
URI 95-06-04 documents inspectors' questions regarding electrical
control circuits which utilize full voltage incandescent position
indicating lamps without current limiting protection. The inspectors
were concerned that failure of the bulb could impact the operability of
the associated safety-related components.
This issue was discussed with personnel from NRR and Region II on
April 20, 1995.
Based on the relatively low probability of occurrence
and the licensee's planned replacement of the susceptible bulbs with
LEDs equipped with an integral resistor, it was determined that no
additional effort is required on this issue. Accordingly, this item is
closed.
RESULTS:
One Non-Cited Violation was identified in this evaluation area which
involved the alteration of test data.
The inspectors reviewed selected engineering issues to determine the
effectiveness of the licensee's controls in identifying, resolving, and
preventing problems.
Included in this review was an assessment of
design control, design and implementation of plant modifications,
engineering and technical support to other organizations, configuration
management, training and staffing, and self-assessment. Other than the
above referenced issue, the engineering program was effectively
implemented.
6.
PLANT SUPPORT (71750)
Safeguards Information Control Concerns
In assessment Report R-SC-95-01, NAD raised concerns on the potential
for compromise of safeguards information while performing word
16
processing of SGI.
This concern was based on the potential that
automatic backups to a PC hard drive were not precluded by plant
procedures.
In response to this concern, the licensee initiated a condition report.
The licensee review performed for the condition report revealed that no
SGI was currently located on PCs or the LAN.
The condition report also
identified corrective actions to preclude this concern in the future.
The inspectors reviewed the condition report.
The inspectors also
performed an independent check of 6 personal computers identified by the
licensee as used to process SGI.
No SGI was found during the
inspectors' review. The inspectors have no further questions on this
issue.
Unsecured Safeguards Information
At approximately 9:00 a.m., on the morning of April 5, 1995, while
performing a routine control room tour, the resident inspectors detected
that certain Safeguard Information was unsecured, stored on a bookshelf
in the shift supervisor's office, which is adjacent to the active
control room. The shift supervisor's office was often un-occupied.
This resulted in the Safeguards Information not being under the control
of an authorized individual. The inspectors brought the matter to the
attention of the site Security Manager. The safeguards information was
moved into the active control room area. The licensee initiated
Condition Report 95-00900 and began an investigation of the situation.
10 CFR 73.21 (d) requires, in part, that while in use, matter containing
Safeguards Information shall be under the control of an authorized
individual. While unattended, Safeguards Information shall be stored in
a locked security storage container.
The licensee's failure to maintain this Safeguards Information under the
control of an authorized individual or in a locked security storage
container is a Violation. VIO 50-261/95-12-05: Failure to Control
Safeguards Information.
RESULTS:
One Violation was identified in this evaluation area concerning the
licensee's failure to maintain control of Safeguards Information.
The inspectors reviewed selected activities of the licensee's programs
for radiological controls, radiological effluents, waste treatment,
environmental monitoring, physical security, emergency preparedness, and
fire protection, to determine if the programs were implemented in
conformance with facility policies and procedures and in compliance with
regulatory requirements. With the exception of the above referenced
safeguards issue, the programs were effectively implemented.
17
7.
EXIT INTERVIEW
The inspectors met with licensee representatives (denoted in
paragraph 1) at the conclusion of the inspection on April 27, 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 p oprietary any of the materials provided to or reviewed by
the inspectors during this inspection. The licensee did not completely
agree with the inspector's characterization of Violation A. The
licensee argued that since the event was self disclosing, the event
review team had recognized the same issues described in the Violation,
and that they were in the process of correcting the identified problems,
that issuing a Violation was capricious and arbitrary. The Senior
Resident Inspector met with top site management after the exit to
elaborate on the substance of the Violation and the guidance offered in
10 CFR 2 Appendix C pertaining to the mitigation of enforcement
sanctions.
Item Number
STATUS
Description/Reference Paragraph
VIO 95-12-01
Opened
Operations Failure to Follow
Procedure During OST-254/paragraph 3
NCV 95-12-02
Opened/Closed
Breaker Clearance Measuring Tools
Not Properly Controlled/paragraph 4
VIO 95-12-03
Opened
Maintenance Planner Fails to
Properly Develop Breaker
PMTR/paragraph 4
NCV 95-12-04
Opened/Closed
Alteration of Test Data/paragraph 5
VIO 95-12-05
Opened
Failure to Control Safeguards
Information/paragraph 6
URI 93-19-03
Closed
Adequacy of Testing for CCW Pump
Autostart Feature/paragraph 4
URI 93-28-03
Closed
Alteration of OST-410
Data/paragraph 5
URI 94-27-09
Closed
EOF/TSC Ventilation
System/paragraph 5
URI 95-06-04
Closed
Potential Design Vulnerability of
Selected Safety-Related Circuits
8.
ACRONYMS AND INITIALISMS
Component Cooling Water
CFR
Code of Federal Regulation
18
Corrective Maintenance
Chemical and Volume Control System
EE
Engineering Evaluation
Engineering Service Request
gpm
Gallons Per Minute
High Efficiency Particulate Absolute
HVH
Heating Ventilation Handling
LAN
Local Area Network
LCO
Limited Condition for Operation
LED
Light Emitting Diode
MDM
Management Designated Monitor
Nuclear Reactor Regulation
OST
Operations Surveillance Test
PC
Personal Computer
Preventive Maintenance
PMTR
Post Maintenance Test Request
Refueling Water Storage Tank
Safeguards Information
Volume Control Tank