ML14181A924
| ML14181A924 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/23/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14181A921 | List: |
| References | |
| 50-261-97-06, 50-261-97-6, NUDOCS 9706050079 | |
| Download: ML14181A924 (19) | |
See also: IR 05000261/1997006
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
50-261
License No:
Report No:
50-261/97-06
Licensee:
Carolina Power & Light (CP&L)
Facility:
H. B. Robinson Unit 2
Location:
3581 West Entrance Road
Hartsville, SC 29550
Dates:
March 23 - April 26, 1997
Inspectors:
B. Desai, Senior Resident Inspector
J. Zeiler, Resident.Inspector
G. Wiseman, Reactor Engineer, Region II
Approved by:
M. Shymlock, Chief, Projects Branch 4
Division of Reactor Projects
Enclosure 2
9706050079 970523
ADOCK 05000261
G
EXECUTIVE SUMMARY
H. B. Robinson Power Plant, Unit 2
NRC Inspection Report 50-261/97-06
This integrated inspection included aspects of licensee operations,
maintenance, engineering, and plant support. The report covers a five-week
period of resident inspection; in addition, it includes the results of an
inspection by one Region II based project engineer.
Operations
Conduct of operations was professional and safety-conscious
(Section 01.1).
Operators properly controlled a 50 percent downpower evolution to repair
a service water leak in the "A" Condensate Pump motor upper oil cooler.
In addition, operators adequately responded to and handled a heater
drain tank level control failure which resulted in a feedwater/
condensate transient. The transient was caused by troubleshooting
activities on the heater drain tank level controller and indicated a
weakness in contingency action planning and understanding of the
potential adverse impact of the activity on the plant (Section 01.2).
Several management changes were announced including Vice President and
Director of Site Operations (Section 06.1).
The Plant Nuclear Safety Committee and the Nuclear Assessment Section
provided strong oversight and safety focus of licensee activities
(Section 07.1).
Maintenance
Routine and emergent maintenance and surveillance activities were
performed satisfactorily. In general, good work control and
coordination was observed, however, several minor problems were
identified indicating a need for greater attention to detail by
maintenance personnel following work instructions and understanding the
work scope prior to implementation (Section M1.1).
A Violation was identified for inadequate Engineered Safeguards (ESF)
testing procedures, in that, they allowed ESF trains to be inoperable
for testing, without invoking a Technical Specification .action statement
(Section M3.1).
A Non-Cited Violation was identified for an inadequate Safety Injection
system check valve test procedure which allowed a testing configuration
that placed the Safety Injection system in a condition outside its
design basis. An Unresolved Item was identified to determine whether
the licensee needed to supplement a Licensee Event Report involving this
issue (Section M8.1).
Engineering
Efforts to strengthen the predictive maintenance program was determined
to be achieving the desired results. A critical self assessment was
performed by the licensee that identified areas of improvement. The
licensee is in the process of developing a program document
(Section E2.1).
Plant Support
A security Hand Geometry system equipment failure was being adequately
investigated. Security personnel failed to properly implement
compensatory measures resulting in an employee whose security badge was
not currently active being allowed to enter the Protected Area (Section
S2.1).
The performance by the fire brigade during a drill was satisfactory and
significantly improved from that observed in October 1996. The use of
additional fire drill props improved fire drill realism and aided the
fire brigade leader in assessing conditions. The fire brigade response
was aggressive and communication and guidance to the off-site fire
department personnel was good (Section F5).
Report Details
Summary of Plant Status
Unit 2 remained at power the entire inspection period. Power was reduced to
approximately 50 percent on April 14 to replace the upper oil cooler to the
"A"
Condensate Pump motor. Power was reduced to approximately 95 percent on
April 17 to assess root cause of minor feedwater oscillations.
I. Operations
01
Conduct of Operations
01.1 General Comments (71707)
The inspector conducted periodic control room tours to verify proper
staffing, operator attentiveness and communications, and adherence to
approved procedures. The inspector attended daily operations turnover,
management review, and plan-of-the-day meetings to maintain awareness of
overall plant operations. Operator logs were reviewed to verify
operational safety and compliance with Technical Specifications (TSs).
Instrumentation, computer indications, and safety system lineups were
periodically reviewed from the Control Room to assess operability.
Frequent plant tours were conducted to observe equipment status and
housekeeping. Condition Reports (CRs) were routinely reviewed to assure
that potential safety concerns and equipment problems were reported and
resolved.
In general, the conduct of operations was professional and safety
conscious. Good plant equipment material conditions and housekeeping
was noted throughout the report period. Specific events and noteworthy
observations are detailed in the sections below.
01.2 Downpower to Repair Condensate Pump Oil Cooler Leak (71707)
a. Inspection Scope
On April 14, 1997, at approximately 1:55 a.m., a downpower to 50 percent
power was initiated to repair a service water leak in the upper oil
cooler to the "A"
Condensate Pump motor.
b. Observations and Findings
The leak was identified by maintenance personnel through the routine oil
sampling program. In order to prevent a motor failure, the licensee
decided to reduce power to remove the pump from service and replace the
oil cooler. Operators properly controlled the downpower and stabilized
the plant at 50 percent power.
Following the downpower, the heater drain pumps were secured to support
troubleshooting of the Heater Drain Tank Level Controller, LC-1530,
which was not controlling level within the optimum range. During the
troubleshooting, the level control signal failed high causing a
feedwater/condensate transient. The operators used abnormal operating
2
procedures effectively to take manual control of heater drain tank and
condenser hotwell level in order to restore plant conditions to normal.
The transient demonstrated, however,,a weakness in contingency action
planning and understanding of the potential adverse impact of the
troubleshooting activity.
c. Conclusions
The inspector concluded that plant operators appropriately controlled
the downpower evolution and adequately responded to the feedwater/
condensate transient. The transient was caused by troubleshooting
activities on the.heater drain tank level controller and indicated a
weakness in contingency action planning and understanding of the
potential adverse impact of the activity on the plant.
06
Operations Organization and Administration
06.1 Management Changes (71707)
a. Inspection Scope
The licensee announced several management changes during this report
period.
b. Observations and Findings
The management changes announced by the licensee included:
John S. Keenan who was promoted to Site Vice President upon departure of
Scotty Hinnant to Brunswick; Dale Young was promoted to be the Director
of Site Operations, and John Boska was selected as the Operations
Manager due to the departure of Bruce Meyer to assume a position at
Harris Nuclear Plant. The licensee plans to name the new Plant Manager
in the near future. Additionally, Talmage Clements replaced John Boska
as the Superintendent for Design Control, and Jim Hendrickson was
promoted to Supervisor for Reactor Systems within the Robinson
Engineering Support Section (RESS).
c. Conclusions
The inspector will followup and update upon the announcement of the new
PlantManager.
07
Quality Assurance In Operations
07.1. Plant Nuclear Safety Committee and Nuclear Assessment Section Oversight
(40500)
a. Inspection Scope
The inspector evaluated certain activities of the Plant Nuclear Safety.
Committee (PNSC) and Nuclear Assessment Section (NAS) to determine
3
whether the onsite review functions were conducted in accordance with TS
and other regulatory requirements.
b. Observations and Findings
The inspector periodically attended PNSC meetings during the report
period. The presentations were thorough and the presenters readily
responded to all questions. The committee members asked probing
questions and were well prepared. The committee members displayed
understanding of the issues and potential risks. Further, the inspector
reviewed NAS audits and concluded that they were appropriately focused
to identify and enhance safety.
c. Conclusions
The inspector concluded that the onsite review functions of the PNSC
were conducted in accordance with TSs. The PNSC meetings attended by
the inspector were well coordinated and meeting topics were thoroughly
discussed and evaluated. NAS continued to provide strong oversight of
licensee activities.
08
Miscellaneous Operations Issues (92901)
08.1 (CLOSED) Licensee Event Report (LER) 50-261/96-003-00, Condition
Prohibited by Technical Specifications Due to Failure to Maintain Shift
Compliment: This LER promulgated the condition described in Non-Cited
Violation (5-261/96-10-01), documented in NRC Inspection Reports 50
261/96-10 and 50-261/96-11, in which a licensed Senior Reactor Operator
stood seven shifts without a current biannual medical examination.
In response to the event, the licensee initiated and committed to
the following corrective actions: revise operations procedures
OMM-001-1, Operations Unit Organization and Administration, and
OMM-001-5, Training and Qualifications, to require monitoring and
reporting of the medical status of all operators and personnel
assigned to the fire brigade in the Operator Hours Tracking Log.
The inspector reviewed the Operator Hours Tracking Log and
verified that the medical qualifications of the April 1997
operations personnel was up-to-date. The inspector reviewed these
completed corrective actions and determined that they were
completed satisfactory. This item is closed.
08.2 (CLOSED) LER 50-261/96-004-00, Manual Initiation of Reactor Protection
System (RPS) due to Turbine Governor Valve Failure: This LER described
the manual reactor trip that occurred on-September 7, 1996. This event
was discussed previously in section 01.4 of NRC Inspection Report
50-261/96-11. The inspector determined that the licensee had adequately
determined the root cause of the trip and corrected the equipment
related failure.
4
The inspector reviewed the licensee's corrective actions which included
the addition of angle mounts and weather covers to the turbine governor
valve actuators to eliminate looseness and broken bolts, soldering and
covering the wire terminations with heat shrink protective shields, and
revisions to relevant maintenance work instructions to caution workers
of the fragile wires in the valves. The inspector determined that
adequate corrective actions were completed. This item is closed.
II. Maintenance
M1
Conduct of Maintenance
M1.1 General Comments (61726 and 62707)
a. Inspection Scope
The inspector observed all or portions of the following maintenance
related Work Requests/Job Orders (WRs/JOs) and surveillances and
reviewed the associated documentation:
WR/JO 97-ABFZ1
Repair Pressurizer Pressure Control
Station,
WR/JO 97-ABKM1
Implement Modification.ESR 97-00198 to
Provide Alternate Reactor Containment Sump
Discharge Flowpath,
WR/JO 97-ABII1
Investigate Tripping of "B"
Drive Motor Generator Breaker,
WR/JO 97-ABGH1/2
Repair Leak on Emergency.Diesel Generator
A Standby Circulating Coolant Pump,
WR/JO 97-ACIR1
Auxiliary Feedwater to "C" Steam Generator
Transmitter FT-1425C Calibration,
OST-902
Containment Fan Coolers Component Test,
Revision 24, and.
Containment Isolation Valves Leakage Test
for WD-1722, Revision 0.
b. Observations and Findings
The inspector observed that these activities were performed by personnel
who were experienced and knowledgeable of their assigned tasks. Work
and surveillance procedures were present at the work location and being
adhered to. Procedures provided sufficient detail and guidance for the
intended activities. Detailed plans were developed with good support
from engineering for troubleshooting activities associated with the
pressurizer pressure control malfunction and control rod drive motor
5
generator trip. Activities were properly authorized and coordinated
with operations prior to start. Test equipment in use was calibrated,
procedure prerequisites were met, system restoration was completed, and
surveillance acceptance criteria were met. Specific observations and
details for several of these activities included the following:
WR/JO 97-ABKM1: This activity was associated with the implementation of
a temporary modification to provide an alternate flowpath from the
reactor containment sump pumps to the Waste Holdup Tanks. The alternate
flowpath involved connecting a pressure rated fire hose from the
discharge vent on the sump pumps to a vent valve in the reactor coolant
drain tank pump suction line which also discharged to the Waste Holdup
Tanks. The inspector observed a minor problem when maintenance
personnel tied off the hose to the incorrect line. A different line for
this purpose had been seismically evaluated and was specified in the
modification implementing procedure. The inspector alerted the engineer
who was monitoring the activity and the discrepancy was corrected.
Since the problem was corrected prior to the actual signoff and
completion of the job, the inspector determined that a procedural
violation had not occurred. The inspector concluded that this problem
was an example of lack of attention to detail on the part of maintenance
personnel in following the work instructions and poor engineering
personnel overview of the activity which allowed the condition to go
uncorrected until questioned by the inspector.
WR/JO 97-ABGH1/2: While mechanical maintenance personnel were
completing the installation of a new circulating water coolant pump and
motor assembly on the "A"
Emergency Diesel Generator, the inspector
noted that the instructions in WR/JO 97-ABGH2 required that
Instrumentation and Control (I&C) perform an electrical resistance check
of the motor following installation. The WR package, however, did not
include a copy of the procedure for performing the motor check and the
I&C technician who was about to start electrical connection of the motor
was not aware that the test was prescribed in the WR instructions.
Following discussions with hi.s supervisor, the I&C technician obtained
the procedure and completed the test. The inspector concluded that this
minor problem demonstrated a need for better understanding of the work
scope by the I&C personnel involved.
c. Conclusions
The inspector concluded that maintenance and surveillance activities
were performed satisfactorily. In general, good work control and
coordination was observed, however, several minor problems were
identified indicating a need for greater attention to detail in
following work instructions and understanding the work scope prior to
implementation.
6
M3
Maintenance Procedures and Documentation
M3.1 Engineered Safeguards Periodic Surveillance (71707, 61726)
a. Inspection Scope
The inspector observed portions of maintenance surveillance test (MST)
procedure MST-23, Safeguards Relay Rack Train "B" (Monthly). MST-23 and
MST-22, Safeguards Relay Rack Train "A" (Monthly) implement the Robinson
TS, Section 4.1, Operational Safety Review, Table 4.1-1, Item 27
requirements to monthly test the Engineered Safeguards Features (ESF)
logic circuitry.
- b. Observation and Findings
The inspector noted that during the performance of MST-22 and MST-23, no
TS action statement was considered applicable by the control room
operating crew. During the performance of MST-22, portions of the
automatic actuation capability associated with ESF Train "A"
was
sequentially bypassed to accommodate logic testing, thus precluding
starting the respective ESF Train "A"
components. The performance of
MST-23 had a similar effect on Train "B".
Upon further discussion, the
inspector was informed that Robinson TS did not specifically prescribe
an action statement during ESF logic testing, and consequently, an
action statement had never been considered to be in effect.
The inspector reviewed the standard Westinghouse TS and confirmed that
it prescribed a very restrictive action statement of two hours to bypass
each ESF train during testing. The restrictive action statement was
based on one train of multiple ESF components being out-of-service
simultaneously. The inspector discussed this with the licensee.
The licensee reviewed the issue and deliberated whether to enter TS 3.0
during the future performance of MST-22 and MST-23. The licensee then
made a decision that TS 3.6.3.d associated with automatic containment
isolation valves would be more appropriate. This TS prescribes a four
hour action statement and was the most restrictive.action statement
affecting ESF components. This effectively allowed four hours to bypass
each ESF trains for testing. A Night Order related to this issue was
issued, in addition to a condition report. The inspector noted that
during the TS conversion to the new improved standard TS this item was
not identified.
c. Conclusions
The inspector concluded that procedures MST-22 and MST-23 were
inadequate, in that, they allowed ESF trains to be inoperable for
testing, without invoking a TS action statement. -The problem
constituted a violation of TS 6.5.1.1.1, Procedures, Tests, and
Experiments. This issue will be documented as Violation (VIO)
50-261/97-06-01: Inadequate Safeguards Procedures that Allowed ESF Train
Being Out-of-Service Without Invoking a TS Action Statement.
7
M8
Miscellaneous Maintenance Issues (92902)
M8.1 (CLOSED) LER 50-261/95-01-00, Safety Iniection Pump Testing Requires TS 3.0 Entry: The licensee identified that the past test configuration for
ASME Section XI Inservice surveillance testing of Safety Injection (SI)
system pump discharge check valves potentially rendered the SI system
inoperable. During partial forward flow testing of the SI pump
discharge check valves (SI-879A B, and C) in accordance with Operations
Surveillance Test (OST) procedure OST-151. Safety Injection System
Components Test, the procedure aligned the normally closed SI test line
to the SI pump discharge header. In this configuration, adequate SI
system flow to the reactor core could not be assured during all design
bases accidents since a portion of the flow would be diverted to the
Refueling Water Storage Tank through the SI test line. As an interim
measure until the issue could be resolved, the licensee entered the
action statement of TS 3.0 on three occasions to complete quarterly
scheduled testing of each of the three SI pump discharge check valves.
The licensee's corrective actions included evaluating the test
configuration to determine if the SI system was actually inoperable with
the SI test line open. The inspector reviewed engineering calculation
RNP-M/MECH-1556, Revision 1, dated April 25, 1995. The results of this
calculation indicated that design SI flow rates would still be
maintained to the Reactor Coolant System (RCS), even with the SI test
line open. However, with a single SI pump in operation, and RCS
pressure near atmospheric conditions, the.increased flow rate would
result in the pump exceeding its runout flow rate. Based on this, the
licensee determined that there was adequate justification for changing
the check valve test frequency from quarterly to a Cold Shutdown and
Refueling Outage interval. The inspector reviewed Technical Management
Manual (TMM) procedure TMM-004, Inservice Inspection Testing, Revision
14, and verified that the change to the testing frequency was properly
documented. In addition, the inspector verified via completed check
valve test data and startup procedures that the valves were being tested
at refueling outage intervals and were scheduled for testing during cold
shutdown conditions. This LER is closed.
The inspector determined that, prior to the licensee recognizing the
adverse test configuration, procedure OST-151 was not appropriate to the
circumstances, constituting a violation of 10 CFR 50, Appendix B,
Criterion V, Procedures. This licensee-identified and corrected
violation is being treated as a Non-Cited Violation (NCV), consistent
with Section VII.B.1 of the NRC Enforcement Policy. This issue will be
documented as NCV 50-261/97-06-02: Inadequate SI Pump Discharge Check
Valve Testing Procedure.
During review of this LER, the inspector noted that a supplement to the
existing LER had not been submitted to the NRC in order to report the
results of the SI operability determination. The past operability
evaluation had concluded that the SI system would have operated at
runout conditions resulting in the inoperability of the system. The
inspector believed that a supplement to the LER, reporting the
8
inoperability, had been required. At the end of the report period, the
licensee had not completed their review to determine whether an LER
supplement was warranted. This issue was identified as Unresolved Item
(URI) 50-261/97-06-03: Review Licensee Assessment of Need to Supplement
LER 50-261/95-001-00.
M8.2 (CLOSED) LER 50-261/96-007-00, Automatic Initiation of RPS Due to Steam
Generator Feedwater Level Control System Failure: On October 20, 1996,
an automatic reactor trip occurred from 20 percent power during startup
from Refueling Outage 17. The reactor trip occurred on Steam Generator
high water level after the "B"
feedwater regulating valve
automatic/manual control station failed to properly control steam
generator level.
During the previous inspection report period covering
the date of this trip, the inspector reviewed the licensee's post-trip
report and plant data 'to ensure that plant equipment operated properly
in response to the trip. With the exception of several minor operator
training enhancements, no problems were identified.
The licensee was unable to determine the exact cause of the controller
failure and concluded that the most likely failure mechanism was
controller aging. The "B"
feedwater automatic/manual control station
was replaced and no further problems were experienced. As part of the
licensee corrective actions, the event was reviewed with all licensed
operators during subsequent routine operator training. Preventive
maintenance procedures were implemented to perform periodic diagnostic
testing of automatic/manual control stations to detect potentially
degrading controllers. The inspector determined that adequate
corrective actions were completed. This LER is closed.
III. Engineering
E2
Engineering Support of Facilities and Equipment
E2.1 Predictive Maintenance (37551)
a. Inspection Scope
The inspector reviewed licensee activities relating to the predictive
maintenance program.
b. Observations and Findings
During this report period, as a result of the predictive maintenance
program, the licensee identified several components that were showing
degradation. Oil analysis from the "A"
Safety Injection pump's outboard
bearing indicated an increasing trend in iron and wear particles. The
pump bearing was replaced. Additionally, the "B"
Spent .Fuel Cooling
Pump was refurbished based on vibration and oil analysis indicating a
negative trend.
9
The licensee has recently employed a supervisor/engineer whose is also
responsible for the coordination of the predictive maintenance program.
Efforts are also under way to develop a predictive maintenance program
guidance/procedure to better coordinate future activities.
c. Conclusions
The inspector concluded that efforts are under way to further strengthen
the predictive maintenance program. A critical self assessment was
performed by the licensee that identified areas for improvement. The
licensee is in the process of developing a program document.
E7
Quality Assurance in Engineering Activities
E7.1 Special UFSAR Review
A recent discovery of a licensee operating their facility in a manner
contrary to the UFSAR description highlighted the need for a special
focused review that compares plant practices, procedures and/or
parameters to the UFSAR descriptions. While performing the inspection
discussed in this report, the inspector reviewed selected portions of
the UFSAR that related to the areas inspected. The inspector verified
that for the select portions of the UFSAR reviewed, the UFSAR wording
was consistent with the observed plant practices, procedures and/or
parameters.
E8
Miscellaneous Engineering Issues (92903)
E8.1 (CLOSED) Violation 50-261/96-10-01, Inadequate Corrective Actions for
Solenoid Operated Valve (SOV) Design Discrepancies: This violation
involved inadequate licensee corrective actions to address configuration
control problems associated with air-operated SOVs being overpressurized
beyond their design limits. In 1988, the licensee failed to adequately
evaluate whether problems existed at Robinson after being alerted to the
potential problem via NRC Information Notice 88-24, Failures of Air
Operated Valves Affecting Safety-Related Systems. Following a
subsequent review of the Information Notice in 1995, the licensee again
failed to adequately evaluate the extent of condition, or determine the
operability impact of 14 safety-related SOVs that were identified with
operating pressure exceeding their design limits.
The licensee responded to the violation by letter dated October 13,
1996. The root cause was attributed to personnel failure to follow
operating experience and corrective action procedures after conditions
adverse to quality were identified. The inspector reviewed these
procedures and determined that they had been enhanced to provide more
specific guidance to plant personnel if potential operability concerns
are identified. In addition, engineering personnel were provided real
time training to emphasize management expectations in this area.
The licensee conducted an evaluation of all safety-related and non
safety-related SOVs to identify and evaluate any SOVs in applications
10
where their design pressure limits were exceeded. The inspector
reviewed the results of this evaluation and determined that it was
comprehensive and detailed. Discrepancies identified by the licensee
were being adequately resolved. This item is closed.
E8.2 (CLOSED) Inspector Followup Item (IFI) 50-261/96-12-03, Review Licensee
Justification for not Completing Non-Validated DBD and GID Evaluations:
This issue was raised during a review of licensee corrective actions for
Violation 50-261/94-23-01 involving the failure to properly revise plant
documents based on the results of a non-validated Containment Isolation
Generic Issue Document (GID) evaluation. Initial licensee corrective
actions for this violation had been to implement a procedure to ensure
that plant personnel perform formal reviews of non-validated GIDs and
Design Basis Documents (DBDs) to .ensure that there were no other plant
documents that needed revision. However, only one non-validated DBD
involving the Main Steam System was.formally reviewed. Since the
results of this review did not identify any documents that needed
revision, the licensee believed it unlikely that there were any other
problems with the other 14 non-validated GIDs and DBDs. Based on this,
the licensee decided not to perform a formal review of the remaining
documents.
The inspector did not consider this adequate justification for not
completing the original scope of the corrective actions for Violation
50-261/94-23-01. During subsequent discussions with the licensee
concerning this matter, the licensee committed to complete the review of
all other non-validated GIDs and DBDs. The inspector reviewed CP&L
letter dated February 11, 1997, which provided the licensee's response
to the NRC's request pursuant to 10 CFR 50.54(f) regarding the adequacy
and availability of design bases information at CP&L nuclear plants. As
part of their enhancements to provide additional assurance that the
plant's design bases had been incorporated into the design, operation,
and maintenance of the plant, the licensee committed to complete the
review of all other non-validated GIDs and DBDs by October 15, 1998.
Based on this commitment, the inspector concluded that this IFI was
closed.
IV. Plant Support
R1
Radiological Protection and Chemistry Controls (71750)
R1.1 Tours of the Radiological Control Area (RCA)
The inspector periodically toured the RCA during the inspection period.
Radiological control practices were observed and discussed with
radiological control personnel including RCA entry and exit, survey
postings, locked high radiation areas, and radiological area material
conditions. The inspector concluded that radiation control practices
were proper.
Si
Conduct of Security and Safeguards Activities (71750)
51.1 General Comments
During the period, the inspector toured the protected area and noted
that the perimeter fence was intact and not compromised by erosion or
disrepair. Isolation zones were maintained on both sides of the barrier
and were free of objects which could shield or conceal an individual.
The inspector periodically observed personnel, packages, and vehicles
entering the protected area and verified that necessary searches,
visitor escorting, and special purpose detectors were used as applicable
prior to entry. Lighting of the perimeter and of the protected area was
acceptable and met illumination requirements.
S2
Status of Security Facilities and Equipment
S2.1 Plant Access Control Equipment Problem and Compensatory Measure
Deficiency (71750)
a. Inspection Scope
The inspector reviewed licensee actions to resolve an operability
problem with the security Hand Geometry system. The inspector also
reviewed the licensee's preliminary investigation into an employee being
allowed to enter the Protected Area (PA) with an inactive security badge
while compensatory measures were being implemented for inoperable hand
geometry equipment.
b. Observations and Findings
On April 6, 1997, the security Hand Geometry system malfunctioned
resulting in the licensee declaring the system inoperable. Compensatory
measures were implemented for controlling personnel entry into the PA.
These measures included assignment of two security personnel at the PA
entry point with post orders to check two forms of personnel
identification for each individual seeking entry into the PA and
verifying each individual was on the authorized access list.
On April 8, while these compensatory measures were still active, an
employee whose security badge was not currently active was allowed to
enter the PA. The employee's badge had been terminated by mistake on
March 27, while the employee was on sick leave. Upon entering the PA,
the employee contacted the Fitness For Duty organization to determine
whether any testing was necessary for returning to work. At this time,
Fitness For Duty personnel identified that the employee's badge had been
terminated. After contacting security, the employee was located and
escorted out of the PA. The employee had been in the PA for
approximately twelve minutes unescorted.
The licensee initiated an investigation of the incident and checked the
access authorization status of all other personnel granted plant access
since implementing the compensatory measures to ensure that no other
12
instances of this type had occurred. No further instances were
identified. The licensee determined that the two security personnel
assigned at the PA entry point failed to verify the individual's access
levels prior to allowing the individual access to the PA. Each of the
two security persons involved in the incident had assumed that the other
person had performed the required access verification indicating a lack
of clear assignment of task responsibility. Following the incident, the
licensee revised the post orders to include specific assignment of the
tasks between security personnel assigned to this post. The inspector
periodically observed implementation of the revised post orders to
verify adequate security personnel performance. No other problems were
identified. The licensee plans to submit an LER for the unauthorized
entry. The inspector will review completion of the licensee's
investigation and corrective actions during review of the LER.
The malfunction in the Hand Geometry system involved an intermittent
problem with the transfer of employee.hand profile data between the
turnstyle hand readers and the main security computer. On April 17, the
problem was temporarily corrected and the system was placed back in
service. The inspector determined that the problem did not involve the
potential for an unauthorized individual to gain access due to hand
geometry reading errors. At the end of the report period, the licensee
was continuing troubleshooting and discussions with the vendor to fully
understand and prevent any future data transfer problems.
c. Conclusions
The inspector concluded that the licensee was adequately investigating
the cause of the Hand Geometry system failure. Security personnel
failed to properly implement compensatory measures for the inoperable
equipment resulting in an employee whose.security badge was not
currently active being allowed to enter the Protected Area. Further
review and enforcement disposition of this incident is planned during
followup of the LER.
F5
Fire Protection Staff Training and Qualification (71750)
F5.1 Fire Drill (71750)
a. Inspection Scope
On April 15, 1997, at 8:00 p.m., the inspector witnessed an announced
backshift fire brigade drill (Fire Drill Number 97-2Q-01-SHIFT-2) with
assistance of the local off-site fire department located in Hartsville,
b. Observations and Findings
The fire drill simulated a lubricating oil spill of approximately 150
gallons with a fire involving two levels of the Unit 2 Turbine Building
(Plant Fire-Area "G" - Fire Zone 25). The simulated fire involved the
Unit 2 turbine lubricating oil reservoir and filter pumps.
13
The inspector observed the drill controllers at the site directing the
fire drill.
The drill was conducted using fire drill planning guide
FBS-01R. The licensee used fire drill props such as tape to simulate
the oil spill area, strobe lights to identify the involved fire areas of
both the lower and second level turbine decks, computer generated
pictures to define the fire status and a device that produced fog for
simulation of smoke.
The control room sounded the plant fire alarm within two minutes of the
report of the simulated fire. The Unit 2 fire brigade team leader and
four fire brigade members responded promptly in full protective clothing
with appropriate fire fighting equipment. Additional personnel from
Unit 2 operations, security and the Hartsville Fire Department, who were
pre-staged onsite, also responded to the drill.
An offensive fire
attack was mounted utilizing two 1/2-inch attack foam fire hose lines
from opposite sides of the fire on the lower level turbine deck,
followed by additional 1/2-inch foam lines extended to the second level
turbine deck. The fire brigade leader properly deployed the fire
brigade personnel, established a command post and effectively used radio
communications. The control room properly used the fire protection Pre
Plans, OMM-003, and simulated tripping the plant due to potential fire
damage. A drill critique was conducted with the fire brigade members
following the drill to discuss the drill, participants performance and
recommendations for improvements. The drill objectives were met.
c. Conclusions
The inspector concluded the drill was performed in a controlled manner
and provided realistic training for the fire brigade. The drill
scenario was challenging. Fire brigade performance has improved.
During observation of a fire drill, the brigade exhibited good command
and control, fire ground tactics, and recovery operations.
V. Management Meetings
X1
Exit Meeting Summary
The inspector presented the inspection results to members of licensee
management at the conclusion of the inspection on May 1, 1997. The
inspector asked the licensee whether any materials examined during the
inspection should be considered proprietary. No proprietary information
was identified.
X3
Management Meeting Sumary
On March 25, 1997, the licensee was presented the results of the NRC
Systematic Assessment of Licensee Performance (SALP) results by the
Region II
Regional Administrator, Luis A. Reyes. The presentation was
open to the public, as well as local officials. The licensee received a
SALP 1 rating in Operations and Plant Support and a SALP 2 rating in
Engineering and Maintenance.
14
PARTIAL LIST OF PERSONS CONTACTED
Licensee
H. Chernoff, Supervisor, Licensing/Regulatory Programs
J. Clements, Manager, Site Support Services
D. Crook, Senior Specialist, Licensing/Regulatory Compliance
J. Keenan, Vice President, Robinson Nuclear Plant
D. Winters, Acting Manager, Operations
G. Miller, Manager, Robinson Engineering Support Services
R. Moore, Manager, Outage Management
D. Stoddard, Manager, Operating Experience Assessment
R. Warden, Manager, Nuclear Assessment Section
T. Wilkerson, Manager, Regulatory Affairs
D. Young, Director, Site Operations
NRC
B. Desai, Senior Resident Inspector
J. Zeiler, Resident Inspector
15
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and
Preventing Problems
IP 61726:
Surveillance Observations
IP 62707:
Maintenance Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 92901:
Followup - Operations
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened.
Ijye Item Number Status
Description and Reference
50-261/97-06-01
Open
Inadequate Safeguards Procedures that
Allowed ESF Train Being Out-of-Service
Without Invoking a TS Action Statement
(Section M3.1)
50-261/97-06-02
Open
Inadequate SI Pump Discharge Check Valve
Testing Procedure (Section M8.1)
50-261/97-06-03
Open
Review Licensee Assessment of Need to
Supplement LER 50-261/95-001-00 (Section
M8.1)
Closed
lype
Item Number Status
Description and Reference
LER
50-261/96-003-00 Closed
Condition Prohibited by Technical
Specifications Due to Failure to Maintain
Shift Compliment (Section 08.1)
LER
50-261/96-004-00 Closed
Manual Initiation of Reactor Protection
System (RPS) due to Turbine Governor Valve
Failure (Section 08.2)
LER
50-261/95-001-00 Closed
Safety Injection Pump Testing Requires TS 3.0 Entry (Section M8.1)
50-261/97-06-02
Closed
Inadequate SI Pump Discharge Check Valve
Testing Procedure (Section M8.1)
16
LER
50-261/96-007-00 Closed
Automatic Initiation of RPS Due to Steam
Generator Feedwater Level Control System
Failure (Section M8.2)
50-261/96-10-01
Closed
Inadequate Corrective Actions for Solenoid
Operated Valve (SOV) Design Discrepancies
(Section E8.1)
IFI
50-261/96-12-03
Closed
Review Licensee Justification for not
Completing Non-Validated DBD and GID
Evaluations (Section E8.2)
Discussed
Type Item Number
Status
Description and Reference
None