ML14178A365
| ML14178A365 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 06/30/1993 |
| From: | Christensen H, William Orders NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A363 | List: |
| References | |
| 50-261-93-11, NUDOCS 9308050101 | |
| Download: ML14178A365 (18) | |
See also: IR 05000261/1993011
Text
rV
REG&z
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
o
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report No.: 50-261/93-11
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: May 15 - June 11, 1993
Lead Inspector: )6De
d
W. T. Orders, Senior Re
ent I pector
Signed
Other Inspectors: L. W. Garner, Senior Resident Inspector
C. R.fgle, Resident Inspector
Approved by:
H. 0. Christensen, Section Chief
Dfte Signed
Division of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection was conducted in the areas of operational
safety verification, surveillance observation, maintenance observation,
meetings with local officials, and followup.
Results:
A violation with two examples was identified involving: failure to follow
procedures regarding a heat trace circuit which remained in alarm for
approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and failure to follow procedures during the performance
of an operations surveillance test when control room operators altered the
chemical volume control system operation and configuration outside the scope
of the procedure without initiating a temporary procedure change -or discussing
their actions with the shift supervisor (paragraph 3).
Another violation was identified for failing to report to the NRC, within 4
hours of notification of South Carolina Department of Health and Environmental
Control authorities, that weir discharge temperature limits had been exceeded.
(paragraph 3).
9308050101 930702
ADOCK 05000261
G
2
A third violation was identified for failure to properly maintain procedure
CM-008, Steam Driven Auxiliary Feedwater Pump, Turbine, and Auxiliaries
Maintenance, in that a precaution to verify the turbine overspeed trip
setpoint following maintenance had been erroneously deleted (paragraph 4).
An unresolved item was identified concerning the potential inoperability of
the boric acid storage tanks involving indications of temperatures less than
the Technical Specification minimum (paragraph 3).
Another unresolved item was identified for the artificial temperature offset
required to force the boric acid heat trace annunciators to alarm at the
nominal setpoints and for the two boric acid heat trace indication circuits
which were discovered to be disabled (paragraph 5).
A third unresolved item was identified concerning the degradation of the A
emergency diesel generator ventilation system involving a damper that had been
manually blocked open (paragraph 3).
A weakness was identified, in that the annunciator panel procedure for a boric
acid heat trace trouble alarm failed to provide actions for a high temperature
condition (paragraph 3).
.Another
weakness was identified, in that the calibration procedures for
reactor coolant system water level instruments failed to include a functional
test of the control room alarm (paragraph 6).
REPORT DETAILS
1.
Persons Contacted
- R. Barnett, Manager, Projects
C. Baucom, Senior Specialist, Regulatory Compliance
D. Bauer, Regulatory Compliance Coordinator, Regulatory Compliance
- 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
S. Farmer, Manager, Engineering Programs, Technical Support
R. Femal, Shift Supervisor, Operations
W. Flanagan Jr., Acting Plant General Manager
- W. Gainey, Manager, Plant Support
- J. Harrison, Manager, Regulatory Compliance
B. Harward, Manager, Engineering Site Support, Nuclear Engineering
Department
P. Jenny, Manager, Emergency Preparedness
D. Knight, Shift Supervisor, Operations
A. McCauley, Manager, Electrical Systems, Technical Support
R. Moore, Shift Operations Coordinator, Operations
D. Morrison, Shift Supervisor, Operations
- C. Olexik, Manager, Plant Assessment Department
A. Padgett, Manager, Environmental and Radiation Control
D. Seagle, Shift Supervisor, Operations
M. Scott, Manager, Performance Engineering, Technical Support
E. Shoemaker, Manager, Mechanical Systems, Technical Support
W. Stover, Shift Supervisor, Operations
- A. Wallace, Acting Operations Manager
- D. Waters, Manager, Regulatory Affairs
- A. Whitehead, Manager Plant Support Services
D. Winters, Shift Supervisor, Operations
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
NRC Management Visits
H. Christensen, Chief, Projects Section lA, Division of Reactor
Projects, visited the site on June 9 and 10, 1993. Mr. Christensen
toured the facility with the residents and attended the meetings with
the local officials. He also met with the licensee's Manager of Project
Management.
- Attended exit interview on June 17, 1993.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2
2.
Plant Status
The unit began the report period operating at full power and continued
operation at or near full capacity until approximately 12:40 p.m., on
the afternoon of May 25. At that time, the licensee initiated a power
decrease to 60 percent to reduce the circulation/service water discharge
temperature to Lake Robinson in order to comply with the NPDES permit.
Unit operation was limited to 60 percent power for the rest of the month
of May. On June 2, following repair of a leaking air fitting on one of
the main feedwater regulating valves and after an increase in allowable
weir discharge limits under the NPDES permit, power was increased to 100
percent. The unit operated at full power for the remainder of the
report period. At the end of this report period, the unit had completed
260 days of continuous operation.
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, Operations 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.
Failure To Make Timely NRC Notification
On the morning of May 25, 1993, the inspectors were advised by the
Manager of Regulatory Compliance of a pending notification to State of
South Carolina authorities for weir discharge temperatures in excess of
NPDES permit limits. During this initial conversation, the inspectors
.II
3
questioned the licensee relative to the need for a 4-hour non-emergency
notification to the NRC in accordance with 10 CFR 50.72 (b)(2)(vi).
Later that day, in response to a licensee request, the inspectors
contacted the NRR project manager responsible for Robinson to obtain
confirmation that an NRC notification should be made if the State was
notified of this event. On the morning of May 26, 1993, the inspectors
relayed the conclusion of the NRR project manager's investigation to the
licensee, which was that a 4-hour non-emergency notification was
required.
At 12:53 p.m., on May 26, 1993, the licensee made an information only
call to the NRC. In that call, the licensee provided information on six
daily maximum weir discharge temperature excursions above the 98.6 *F
NPDES permit limit. These excursions occurred from May 15, 1993, to
May 20, 1993. Additionally, the licensee also provided information on
compensatory measures taken as a result of the elevated weir discharge
temperatures. These measures included reducing Unit 2 power to 60
percent and taking Unit 1 off line. During subsequent discussions with
the inspector, the licensee indicated that the original intention was to
transmit the information to the HDO to satisfy the requirements of
10 CFR 50.72 (b)(2)(vi). However, after discussions with the HDO and in
the absence of knowledge that formal written notification to the State
had been made from the licensee's corporate office, the report was
communicated to the NRC as "info only".
Following this notification, the inspectors requested clarification
from the licensee of information provided from CP&L corporate
(Environmental Services) to South Carolina state officials. As a result
of this request, site personnel determined that a letter detailing the
temperature excursions in excess of the NPDES permit limits had been
transmitted from CP&L corporate offices to South Carolina state
officials on the evening of May 25, 1993. Accordingly, at 4:45 p.m., on
May 26, 1993, the licensee made a 4-hour non-emergency notification to
the NRC in accordance with 10 CFR 50.72 (b)(2)(vi) documenting this
correspondence. The failure to make a timely notification to the NRC
documenting notification of State authorities of excessive weir
discharge temperatures is considered a violation. VIO: 93-11-01, Failure
To Make A Timely Notification To The NRC Of A Notification To State
Authorities.
On June 2, 1993, CP&L Corporate Environmental Services advised the State
of South Carolina that the weir discharge limit for the month of May was
exceeded. This was followed by a 4-hour non-emergency notification in
accordance with 10 CFR 50.72 (b)(2)(vi). The inspectors determined that
this notification was made within the required time frame.
Failure To Follow Procedures
[Boric Acid Heat Trace Circuit 1 In Alarm]
At 8:30 a.m., on May 20, 1993, the inspectors noted that the local
annunciator for heat trace circuit I was in alarm, indicating a high
4
temperature situation of approximately 207' F. This circuit provides
heat trace protection for the suction and discharge piping associated
with BATP B between valves CVC-379, CVC-341, CVC-334, and CVC-336.
After the inspectors questioned the operator about the alarm, the
associated piping was cooled and the alarm cleared by recirculating the
contents of BAST B. Following the recirculation, the temperature of
the circuit again rose to greater than 200* F. A subsequent adjustment
to the circuit's thermostat by I&C personnel restored heat trace circuit
1 to normal operation.
In response to the inspectors' questions, the on-shift, inside AO stated
he was aware that the circuit was in alarm, but had been unable to
investigate the cause of the alarm since turnover (about 90 minutes
before). The inspectors reviewed the strip chart recording of the
circuit's temperatures and determined that it had been at or near the
alarm setpoint since 2:45 a.m. that morning. The strip chart recording
had been reviewed and initialed 4 times by two different operators
during the 6-hour period the high temperature alarm existed before the
condition was questioned by the inspectors. The previous shift inside
AO and shift supervisor stated that they were both aware of the alarm
condition. However, they erroneously attributed the alarm to residual
heat from the motor/pump generated from the performance of OST-107,
Boric Acid Blender Control, Valve and Pump Operation, which had been
conducted earlier on their shift. The previous shift AO and shift
supervisor also stated that the temperature of the circuit declined near
the end of the shift. From the stripchart, the inspectors determined
that the temperature of the circuit did decrease approximately 2 degrees
from about 3:30 a.m. to 7:00 a.m., on May 20, 1993. However, a degree
of this change appeared to be an offset which occurred coincident with
the change in the stripchart paper at 6:20 a.m. The prior shift inside
AO and shift supervisor acknowledged that they had taken no action to
determine the cause of the alarm or to reduce the temperature of heat
trace circuit 1 after the high temperature condition occurred.
Furthermore, the inspectors determined that the condition was not noted
in either the AO's logs or the shift supervisor's logs. Additionally,
no turnover was conducted on this item with the oncoming shift.
The inspectors reviewed the guidance available to Operations personnel
for alarms on boric acid heat trace circuits. A portion of this
guidance was in the form of an E-Mail memo to the shifts from the Shift
Operations Coordinator dated April 19, 1993. That guidance required
that inside A~s evaluate and initial heat trace chart recordings every 2
hours. This E-mail also discussed applying the appropriate LCO for TS
circuits in alarm. However, the E-mail focused on heat trace circuit
alarms for low temperature. Annunciator procedure APP-036-H2 provided
operator actions for Boric Acid Heat Trace Trouble Alarms on panel APP
036. Step 2 of this procedure requires the operator to determine the
reason for the abnormality and gives examples of deficiencies which
could result in alarms on heat trace circuits. As in the E-Mail memo,
the majority of actions of APP-036-H2 are directed at resolving low
temperature conditions.
5
OMM-001, Operations-Conduct of Operations, Section 5.9.2, Annunciator
Panel Procedure Guidelines, requires that "when the diagnosis of the
alarm concludes that the actions listed in the APP are not appropriate,
then the existing plant conditions, diagnosis conclusion, and the
actions taken shall be logged." Additionally, OMM-001, Section 5.16,
Local Panel Indication, states: "Corrective action should be initiated
on local controls and indications to ensure proper system operation."
The failure of Operations personnel to determine the cause of the alarm
as required by APP-036-H2; to document their conclusions regarding the
alarm as required by OMM-001;
and to take corrective actions as
specified by OMM-001 constitutes a violation for failure to determine
cause, log the diagnosis of, and take corrective actions for the heat
trace circuit 1 alarm condition. This event denotes one of two examples
which collectively comprise VIO: 93-11-02, Failure To Follow Procedures
For a Heat Trace Circuit In Alarm/Failure To Follow Procedure During
Performance Of OST-254.
As discussed above, APP-036-H2 primarily addressed low temperature alarm
conditions on boric acid heat trace circuits. The failure of APP-036-H2
to provide actions for a high temperature condition is considered a
weakness.
[Residual Heat Removal System Leak Check]
On May 27, 1993, while performing a routine leak test surveillance of
the RHR system, control room operators determined that valve HCV-142
(RHR To Letdown Line Isolation Valve) would not open after they had
adjusted letdown pressure to 350 psig as required by applicable
procedure OST-254, Residual Heat Removal System and RHR Loop Sampling
System Leak Test. The air operated valve was visually examined and it
was found that although it had a full open signal, as evidenced by local
air pressure indications, the valve was closed. The operators discussed
the issue amongst themselves and concluded that the valve's failure to
open was probably attributed to the 350 psig pressure differential
across it. Accordingly, after they referred to steam tables, the
operators concluded that letdown pressure could be reduced to 104 psig
without flashing the system. They theorized that this lower pressure
would allow the valve to open.
Without consulting the shift supervisor or processing a temporary
procedure change, the operators reduced letdown pressure until valve
HCV-142 opened. The pressure was then returned to 350 psig and that
portion of the test completed. The actions taken to reduce pressure to
accomplish the test were then discussed with the shift supervisor. At
the shift supervisor's direction, a temporary change to the procedure
was developed and the test completed.
This act not only constituted a failure to follow procedure OST-254, but
was in violation of Technical Specification 6.1.1.5 which specifies the
requirements for making temporary procedure changes.
6
Technical specification 6.5.1.1, Procedures, Tests and Experiments,
requires in part that written procedures be established, implemented,
and maintained concerning the activities delineated in Appendix A of
Regulatory Guide 1.33, Rev. 2, February 1978, which in turn specifies in
parts 1.d, 3.d, and 3.n procedures for procedure adherence, temporary
procedure changes, operation of the emergency core cooling system, and
operation of the chemical and volume control system, respectively.
Administrative procedure AP-006, Procedure Adherence, states in section
5.1 that adherence to approved plant operating procedures is mandatory.
Section 5.2 delineates the only three mechanisms through which a
deviation from an approved procedure can occur; those being either a
permanent procedure change, a temporary procedure change or in an
emergency situation, a "deviation." In this particular situation, in as
much as the evolution was not an emergency situation, a temporary
procedure change appears to have been appropriate.
Technical specification 6.5.1.1.5 requires that temporary changes to
procedures, tests or experiments be approved by two members of the plant
staff, at least one of whom holds a Senior Reactor Operator License.
The temporary change must be documented and reviewed within 21 days to
determine if the change constitutes an unreviewed safety question.
Contrary to the above, on May 27, 1993, control room operators did not
follow OST-254, in that when they were unsuccessful in getting valve
HCV-142 to open when performing step 16 of section 7.1, they altered
CVCS operation and configuration in an attempt to open the valve. This
was done outside the scope of the applicable procedure and without
initiating a temporary procedure change or discussing it with the shift
supervisor.
The above event constitutes the second example of failure to follow
procedures identified in this report which collectively comprise
Violation 93-11-02, Failure To Follow Procedures For a Heat Trace
Circuit In Alarm/Failure To Follow Procedure During Performance Of OST
254.
Feedwater Regulating Valve Air Line Leak
On May 31, 1993, with the Unit at 60 percent power, a steam flow-feed
flow mismatch occurred as a result of the C feedwater regulating valve
(FCV-498) beginning to fail closed. In response to this annunciated
alarm condition, the operators took remote manual control of the valve
in accordance with AOP-010, Inadequate Feedwater Flow. Following
repairs to stop a leak on an air line to the valve's positioner, the
valve was restored to automatic control on June 1, 1993.
The inspectors reviewed a stripchart of steam generator parameters
recorded during the transient and discussed the transient with
operations personnel. The inspector concluded that prompt operator
actions probably minimized the consequences of the transient.
7
The inspectors witnessed portions of the troubleshooting and repair
efforts performed to return the valve to service. Specifically, the
inspectors witnessed a torque check of the C feedwater regulating valve
lock bar capscrews in accordance with WR/JO 93HTY008. This check was
accomplished without incident. In addition, the inspectors were present
for the tightening of a plug on a tee connection in the air line to the
valve's positioner (accomplished per WR/JO 93AFZL). The inspectors also
attended the pre-job brief conducted for this evolution and witnessed
control room activities coincident with the repair. The inspectors
concluded that the appropriate precautions were taken and that the
evolution was well-conducted.
As a part of the post-maintenance inspection efforts, the inspectors
reviewed a safety analysis dated October 18, 1987, approving the
practice of placing a feedwater regulating valve in local manual
control; thereby, defeating a feedline isolation signal for the valve.
This same approach was used for the repairs on June 1, 1993, to minimize
the possibility of a plant transient as a result of the repair efforts.
The conclusion of the analysis was based on a review of the steam break
analysis assuming that the feedwater regulating valve fails to close.
Isolation of the feedline is provided by the feedwater header section
valves. The inspectors have no further questions on the transient or the
repair efforts.
Boric Acid Storage Tank Inoperability
On Monday, May 31, 1993, an operator noted low temperature alarms for
boric acid heat trace circuitry coincident with the recirculation of
both the A and B BASTs. Low temperature alarms were received for heat
trace circuits 3 (BATP B discharge), 35 (BAST B discharge), 7 (piping
downstream of BAST B recirculation valve HCV 105), and 8 (piping
downstream of BAST A recirculation valve HCV 110). The fact that these
low recirculation piping temperatures could indicate temperatures less
than the TS minimum for the BASTs.was recognized by the operators. The
observed heat trace circuit performance and the concerns on BAST
operability were communicated to the system engineer by the operators in
an E-mail message dated May 31, 1993. On Tuesday, June 1, 1993, the
inspectors questioned Operations management on the need for an
operability determination based on the potential for BAST temperatures
of less than 145'F. At 5:00 p.m. that afternoon, the licensee entered a
72-hour operability determination in accordance with OMM-039,
Operability Determination. The determination was completed at 4:00 p.m.
on June 2, 1993, and concluded that both trains of the boric acid
flowpath had been inoperable for an unknown period of time. The
determination also concluded that the recirculation restored the tanks
to an operable condition. It was also noted in the operability
determination that recirculation at an interval of less than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
would prevent the low temperature excursions during recirculation of the
BASTs. Following the operability determination, the licensee instituted
a 4-hour recirculation schedule, with temperatures monitored by
operations personnel, for the in-service BAST A. This tank was
sufficient to satisfy TS requirements for boric acid inventory. At this
8
recirculation frequency, no additional temperature excursions below 145*
were observed in the in-service BAST A recirculation flowpath. The
recirculation frequency of BAST B was varied, while observing
temperatures in the recirculation flowpath, in an effort to extend the
recirculation interval and determine the cause of the temperature
excursions. At the end of the inspection period, the licensee was
recirculating BAST B every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Temperatures of the piping in the recirculation path for BASTS A and B
were observed by ETS personnel during recirculation on June 2, 1993.
Data was collected to determine if temperatures less than 1450 occurred
in any leg of the recirculation piping. The inspectors witnessed the
data gathering, observed temperatures on the heat trace strip chart
recorder, and independently reviewed plots of the observed temperatures.
The minimum temperatures indicated on any segment were 1420F for the A
flowpath and 138*F for the B flowpath. This occurred with the BASTs
indicating approximately 163'F. No temperatures were observed on any
segment of piping below the precipitation point for the existing
concentrations in the BASTs. When the recirculations were repeated 5
hours later, no temperature in either flowpath dropped below the 145*F
TS limit for the BASTs.
At the end of the inspection period, the licensee was attempting to
address the issue of past inoperability of the boric acid system.
Pending the completion of the licensee's operability determination, this
item remains unresolved. URI: 93-11-03, Potential Inoperability Of The
Boric Acid Storage Tanks.
Inadequate Engineering Analysis
On June 5, 1993, an engineering supervisor provided the inspectors with
a written analysis of a B BAST pump performance test performed on
June 4, 1993. The test had been performed to address questions
regarding the potential for boric acid precipitate buildup in the BAST
discharge lines during previous low temperature conditions in the BAST.
During the review of this material, the inspectors noted that the
percentage change of the measured pump discharge pressure from the
manufacturer's pump curve had been used to adjust the measured pump
discharge flow to determine the pump flow rate at design conditions.
However, since the pump curve in the region of interest indicated that
each foot change in pump head would result in more than a 2 gpm increase
in flow, the percentage change in head was not the correct-factor to
utilize in projecting the pump flow at design conditions. The
evaluation of the test data was technically incorrect. This was
discussed with the engineering supervisor. Later, the inspectors were
presented with a technically acceptable analysis; however, the results
of this revised analysis could not be used to conclusively demonstrate
that precipitation had not occurred. At the end of the report period,
the licensee was evaluating other information which would address the
question. The issue of temperatures below the precipitation point for
the boric acid solution will be resolved in the inspectors' review of
URI 93-11-03.
1.
9
Degraded Diesel Generator Ventilation System
On June 10, 1993, during the performance of bi-weekly surveillance test
OST-401 on the A EDG, licensee personnel observed that the room
ventilation air return damper was partially open when it was to have
been closed. The damper is designed to automatically close when the
ambient air temperature is above 55'F. With the damper partially open,
the efficiency of the ventilation system was degraded. The licensee
found that the damper had been manually blocked open by a wooden wedge
which had apparently been manufactured for the purpose. The licensee is
currently evaluating the impact of this illicit modification on the
system's ability to perform its intended safety function. Pending the
completion of that evaluation, this issue will remain unresolved.
URI: 93-11-04, Degraded Diesel Generator Ventilation System.
Two violations were identified.
4. 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 procedure addressed 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, and that
the test conformed to TS requirements.
Specifically, the inspectors witnessed/reviewed portions of OST-202,
Steam Driven Auxiliary Feedwater System Component Test (Monthly).
During post-OST discussions with Engineering Tech Support personnel, the
requirement to test the turbine overspeed trip mechanism following
certain maintenance activities was discussed. Based on the results of
these discussions, the inspectors reviewed ONS Action Item Number 90
010. This item specified that CM-008 (Steam Driven Auxiliary Feedwater
Pump, Turbine, and Auxiliaries Maintenance) be revised to include a
precaution to test the turbine overspeed trip setpoint following renewal
of the trip striker pin or striker spring. This ONS action item was
derived from SOER 89-01. The action item was accomplished through
Revision 8 to CM-008, on August 23, 1990. However, the inspectors noted
that the precaution was not contained in the current revision of CM-008.
Based on an interview of the Manager of Maintenance Support Programs,
the inspectors learned that the precaution was erroneously deleted from
the procedure during Revision 12 to CM-008. This revision became
effective on November 25, 1992, and was accomplished as part of the
procedure upgrade program. The fact that the precaution was the result
of an ONS action item was not clear since Revision 8 to CM-008 failed to
reference this commitment as required by AP-004, Procedure Control.
After the erroneous deletion was identified, the licensee revised CM-008
to include the appropriate precaution on testing the turbine overspeed
10
device and promptly initiated a review of all safety-related maintenance
procedures to ensure their accuracy.
The licensee concluded from their investigation of this error that no
maintenance had been performed on the turbine overspeed mechanism since
Revision 12 to CM-008 had been made. The inspectors independently
reviewed a computerized listing and brief description of all maintenance
performed on the AFW system since 1991 and concurred with the licensee's
observation. Thus, the inspectors concluded that the safety signifi
cance of the deletion was minimal.
This event is similar in nature to an event addressed in NCV 93-05-03.
In that particular situation, a procedure step which had been implement
ed in response to a Violation, was erroneously deleted during a later
revision to the procedure. The corrective actions associated with NCV
93-05-03 included a review of all safety-related maintenance procedures
to ensure that procedure revisions which had been made pursuant to
commitments or corrective actions taken as a result of previously
identified findings, were still embodied in the procedures. The
corrective actions taken at that time did not, however, include a review
of those procedures to verify that previous procedure revisions
implemented to include external inputs such as derived from industry
experience, SOERs, IENs, etc., were still incorporated. Indeed, as
referenced above, the requirement to test the turbine overspeed trip
setpoint following renewal of the trip striker pin or striker spring had
its genesis in SOER 89-01. Accordingly, the licensee's review failed to
detect the erroneous deletion.
Technical Specification 6.5.1.1.1.a. requires that procedures be
maintained for activities referenced in Appendix A of Regulatory Guide
1.33, Revision 2, February 1978. Appendix A, Item 9 requires written
procedures for maintenance that can affect the performance of safety
related equipment. CM-008 provides maintenance instructions for the
steam driven auxiliary feedwater pump. During revision 12 to CM-008, the
precaution to test the overspeed mechanism following renewal of the trip
striker pin or striker spring was erroneously deleted. The failure to
properly maintain CM-008 is a violation. VIO: 93-11-05, Failure To
Properly Maintain Maintenance Procedure CM-008.
During the course of this inspection, the inspectors noted that Opera
tions Surveillance Test procedure OST-202, Steam Driven Auxiliary
Feedwater System Component Test, required that the turbine mechanical
overspeed trip device be exercised before the pump was run for the
test. The inspectors were concerned that this practice constituted pre
dispositioning of the equipment, similar to stroking a valve prior to
performing a valve stroke test. After discussing this issue with the
licensee, they initiated a review of the history associated with
performing the mechanical overspeed trip mechanism test, and elected to
revise the procedure to require the test be performed once per refueling
cycle. The licensee also stated that they would run the pump before and
after testing the trip mechanism to ensure pump operability.
One violation was identified.
5. 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, and
radiological controls were adhered to. In particular, the inspectors
observed/reviewed the following maintenance activities:
WR/JO 93HTY008
Perform Feedwater Regulating Valve Lock
Bar Capscrew Torque Check (FCV-498)
WR/JO 93AFZL1
Troubleshoot FCV-498 To Determine Cause Of
Not Responding Properly In Auto
WR/JO 93BUY341
Calibrate B Boric Acid Storage Tank
Temperature Indicating Controller
WR/JO 93BWR231
Calibrate The L&N Heat Trace Recorders
Troubleshooting Of FCV-498
The inspectors witnessed repair efforts to feedwater regulating valve C
accomplished under WR/JO 93HTY008 and WR/JO 93AFLZ1.
These efforts are
discussed in paragraph 3.
Calibration of Boric Acid Heat Trace Recorder 1
On June 2, 1993, the inspectors witnessed calibration of heat trace
recorder number 1 in accordance with WR/JO 93BWR231. This consisted of
a calibration of the stripchart recorder and the associated
annunciators. The effort was conducted as part of the licensee's
investigation into the observed low temperatures in boric acid
recirculation piping while recirculating the BASTs. During the
calibration of the annunciators, the inspectors noted that the desired
alarm setpoints, as entered into the electronic controller for the
stripchart recorder, were approximately 9 degrees higher than the
nominal 150'F setpoint. This bias was required in order to have the
annunciator alarm when the stripchart recorder indicated approximately
150'F. When questioned by the inspectors, the technicians where unable
to provide any explanation as to why this offset was necessary.
12
Additionally, during the calibration, it was noted that Circuit 31 (the
normal boration line between FT-113 and CVC-354) was not indicating.
Investigation by the technician revealed that the indication for this
circuit had been disabled with a dip switch in the stripchart recorder
housing. During subsequent discussions, the system engineer stated that
he had found another circuit similarly disabled the prior week.
Pending the licensee's resolution of these issues, they remain
unresolved. URI: 93-11-06, Offset Required In Heat Trace Alarm
Circuitry And Disabled Indication For Boric Acid Heat Trace
Temperatures.
No violations or deviations were identified. Except as noted above, the
area/program was adequately implemented.
6. Followup (92700, 92701, 92702)
(Open) LER 92-11, Conditions Outside Design Basis Due to Inadequate
Seismic Restraints. This item, regarding inadequately supported Copes
Vulcan valves, was discussed in IR 92-16. The inspectors verified,
prior to cycle 15, that the supports identified in the LER as not
meeting short-term criteria were modified as committed. The LER
indicated that short-term qualified supports would be upgraded to long
term criteria by no later that RO-15. Design activities were in
progress to develop modifications to address these supports. Based upon
preliminary design activities, there were an estimated 62 supports that
required modifications. Furthermore, there were 38 long-term qualified
supports that were identified to be inspected because as-built
documentation could not be found for some specific attributes.
In addition to the supports associated with the Copes Vulcan valves
discussed above, there were two other processes which have identified
short-term qualified supports. The piping improvement program
identified 109 supports on 8 lines that were short-term qualified and
another 70 supports on various lines that require inspection as describe
above. Normal operating activities (i.e., deficiency identification
program and engineering activities) resulted in the identification of an
estimated 13 additional supports that require upgrade to long-term
criteria and 3 supports that require inspections. Specifically, 4
issues were identified via normal operating activities. These were:
(1) main steam line support embedded plate had broken grout; (2) control
room cooler hot-gas bypass line temperature was outside design
specification; (3) SI and CS pumps suction piping associated with the
recirculation pathway were designed to 100* F verses the 210'F
anticipated after an accident; and (4) the CS headers inside the CV were
analyzed empty, whereas due to the RWST head these headers are partially
filled.
In summary, 184 supports may require upgrading to long-term criteria and
another 111 supports need inspections. Design modification packages
have been completed for 26 supports and 15 supports have been inspected.
Support design was scheduled to be completed by the end of June;
13
however, construction reviews and outage scheduling will not be
completed until August 1, 1993. Thus, although it was anticipated that
the work would be performed before the end of RO-15, there was a
possibility some of the work will not be completed by that time. This
item will remain open pending completion of modification package
development and implementation.
(Closed) VIO 92-27-02, Failure To Implement GP-008 In That RCS Water
Level Instrumentation Loops Were Not Calibrated As Required.
The
inspectors verified that GP-008, Draining The Reactor Coolant System,
had been revised, as stated in the Reply To A Notice Of Violation, dated
December 4, 1992, to require individual sign-offs for calibration of
each component in LT-403 and LT-404 RCS level instrumentation loops.
The inspectors also confirmed that the "Master WR List" for GP-008 was
changed to include calibration of LT-403 and LT-404.
The inspectors noted that the calibration procedure for the alarm
switches LSL-403 and LSL-404 did not include a verification that when a
switch actuates a visible and audible indication is received in the
control room. In addition, through discussions with I&C planners and
document reviewers, the inspectors determined that there were no
procedural requirements to verify that these alarm features work
properly. GL 88-17, Loss Of Decay Heat Removal, enhancement item 1.d
recommended that visible and audible indications of abnormal conditions
in level be provided. The licensee's response to Generic Letter 88-17,
dated February 1, 1989, stated that "An alarm will be added in the
control room for each level indicator."
LSL-403 and LSL-404 were
provided to meet this commitment. Implicit in the recommendation to
provide an alarm is the expectation that the alarm features will be
functionally tested periodically. Failure to verify that the alarm is
received in the control room when these switches are calibrated/tested
is a weakness. The inspectors also noted that since annunciators are
considered to be nonsafety-related, non-MST calibration procedures
typically do not verify proper operation of the alarms/annunciators.
This was discussed with the Maintenance Manager, who disagreed with the
inspectors characterization of this problem as being a weakness.
However, he indicated that procedures for LSL-403 and LSL-404 would be
revised to include alarm verification. The MSTs that were upgraded as
part of the maintenance procedure rewrite program verified that alarms
were received during testing.
(Closed) VIO 92-34-04, Failure To Adequately Establish EPP-9 In That
Steps Were Provided To Operate The RHR Pump At Pressures Above Its
Shutoff Head Without A Caution Note Prior To Steps. The inspectors
reviewed the Reply To A Notice Of Violation dated February 11, 1993.
AP-022, Document Change Procedure, was verified to have been revised as
committed. In addition, the inspectors verified that EPP-9, Transfer To
Cold Leg Recirculation, Revision 11, issued on January 5, 1993, provided
a caution note to warn operators that operation of RHR Pumps in a dead
headed condition for greater than 9 minutes could result in pump damage.
14
The applicability of this violation to other EPPs will be inspected as
part of the followup activities associated with IFI 91-22-10 regarding
the AOP/EOP upgrade project. This item is considered closed.
No violations or deviations were identified. Except as noted above, the
area/program was adequately implemented.
7.
Information Meeting With Local Officials (94600)
On June 9 and 10, 1993, information meetings were held with local
officials representing the City of Hartsville, the County of Darlington
and the County of Florence. The meetings were held in the offices of
the officials at their request. The mission and functional organization
of the NRC and its relationship to the community were discussed. The
meetings were informal with the attendees responding with questions of
interest and importance to their communities. Plant operational safety,
security, emergency plans, and community staffed plant fire response
were of high interest to the groups. The response by the officials was
very positive and provided a good opportunity for interface and followup
communication. H. Christensen, Chief, Project Section IA, Division of
Reactor Projects, attended and participated in the meetings.
8.
Exit Interview (71701)
The inspection scope and findings were summarized on June 17, 1993, 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
93-11-01
VIO:
Failure To Make A Timely Notification To
The NRC Of A Notification To State Authorities
(Paragraph 3).
93-11-02
VIO:
Failure To Follow Procedures For a Heat
Trace Circuit In Alarm/Failure To Follow
Procedure During Performance Of OST-254
(Paragraph 3).
93-11-05
VIO:
Failure To Properly Maintain Maintenance
Procedure CM-008 (Paragraph 4).
93-11-03
URI:
Potential Inoperability Of The Boric Acid
Storage Tanks (Paragraph 3).
93-11-04
URI:
Degraded Diesel Generator Ventilation
System (Paragraph 3).
15
93-11-06
URI: Offset Required In Heat Trace Alarm
Circuitry And Disabled Indication For Boric Acid
Heat Trace Temperature (Paragraph 5).
9.
List of Acronyms and Initialisms
a.m.
Ante Meridiem
Auxiliary Operator
Abnormal Operating Procedure
APP
Annunciator Panel Procedure
BAST
Boric Acid Storage Tank
BATP
Boric Acid Transfer Pump
Component Cooling Water
CFR
Code of Federal Regulations
Corrective Maintenance
Carolina Power & Light
Chemical & Volume Control System
Emergency Operation Procedure
End Path Procedure
Engineering Technical Support
F
Fahrenheit
Flow Control Valve
FT
Flow Transmitter
gpm
Gallons Per Minute
GL
Generic Letter
General Procedure
Hand Control Valve
HDO
NRC Headquarters Duty Officer
Instrumentation & Control
i.e.
That is
IR
Inspection Report
LCO
Limiting Condition for Operation
LER
Licensee Event Report
LT
Level Transmitter
Maintenance Surveillance Test
National Pollutant Discharge Elimination System
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
OMM
Operations Management Manual
Onsite Nuclear Safety
OST
Operations Surveillance Test
p.m.
Post Meridiem
psig
Pounds Per Square Inch - Gage
Refueling Outage
Refueling Water Storage Tank
Safety Injection
Significant Operating Experience Report
TS
Technical Specification
16
Unresolved Item
Violation
W/R
Work Request
WR/JO
Work Request/Job Order