ML14178A355
| ML14178A355 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 06/08/1993 |
| From: | Christensen H, Garner L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A353 | List: |
| References | |
| 50-261-93-10, NUDOCS 9307020169 | |
| Download: ML14178A355 (13) | |
See also: IR 05000261/1993010
Text
p5 V%
REG&y
UNITED STATES
0
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report No.:
50-261/93-10
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: April 17 - May 14, 1993
Lead Inspector:
41
7
9>
7127
L. W. Garner, Senior Resident Inspector
at Signed
Other Inspector:/. R Ogle, Resident Inspector
Approved ky: 471 -i
3
C ristensen, Chief
at Signed
Reacto Projects Section 1A
Division of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection was conducted in the areas of operational
safety verification, surveillance observation, maintenance observation, loss
- of decay heat removal, and followup.
Results:
A violation was identified for failure to prescribe appropriate procedures
to verify the proper operation of the AMSAC A microprocessor after its
replacement (paragraph 5).
A non-cited violation was identified, in that, the provisions of Technical
Specification 6.2.3.b regarding authorization of-shift-work hours in excess of
those specified was not implemented (paragraph 6).
An inspector followup item was identified concerning spare part availability
for the Anticipated Transient Without Scram Mitigation Actuation Circuitry
System (paragraph 5).
9307020169 930608
ADOCK 05000261
Q
2
Greater than 10 gpm Reactor Coolant System leakage from a charging pump
relief valve resulted in a Notice Of Unusual Event emergency classification
(paragraph 6).
While inspecting the inside of a emergency diesel generator panel, the
inspectors observed that a capacitor in the voltage regulator circuit was
damaged (paragraph 3).
REPORT DETAILS
1.
Persons Contacted
- D. Bauer, Regulatory Compliance Coordinator, 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, Robinson Nuclear Project
- W. Gainey, Manager, Plant Support
- H. Habermeyer, Vice President, Nuclear Services
- J. Harrison, Manager, Regulatory Compliance
- P. Jenny, Manager, Emergency Preparedness
D. Knight, Shift Supervisor, Operations
- A. McCauley, Manager - Electrical Systems, Technical Support
R. Moore, Acting Manager - Shift Operations, Operations
D. Morrison, Shift Supervisor, Operations
- P. Musser, Manager - Engineering/Technical Support, Nuclear Assessment
Unit
- A. Padgett, Manager, Environmental and Radiation Control
E. Shoemaker, Manager, Mechanical Systems, Technical Support
W. Stover, Shift Supervisor, Operations
- D. Taylor, Manager, Materials &-Contract Services
- A. Wallace, Acting Manager, Operations
- L. Williams, Manager, Security
D. Winters, Shift Supervisor, Operations
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
- Attended exit interview on May 24, 1993.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Plant Status
Except for a power reduction to perform turbine generator valve testing,
the unit operated at full power during the report period. RCS leakage
greater-than 10 gpm from the A charging pump stabilizer relief valve
resulted in the unit being in a NOUE emergency classification for
approximately four and one-half hours on May 12, 1993. See paragraph 3
for additional information.
2
3. Operational Safety Verification (71707)
The inspectors evaluated licensee activities to confirm that the
facility was being operated safely and in conformance with regulatory
requirements. These activities were confirmed by direct observation,
facility tours, interviews and discussions with licensee personnel and
management, verification of safety system status, and review of facility
records.
To verify equipment operability and compliance with TS, the inspectors
reviewed shift logs, Operation's records, data sheets, instrument
traces, and records of equipment malfunctions. Through work
observations and discussions with Operations staff members, the
inspectors verified the staff was knowledgeable of plant conditions,
responded properly to alarms, adhered to procedures and applicable
administrative controls, cognizant of in-progress surveillance and
maintenance activities, and aware of inoperable equipment status. The
inspectors performed channel verifications and reviewed component status
and safety-related parameters to verify conformance with TS. Shift
changes were routinely observed, verifying that system status continuity
was maintained and that proper control room staffing existed. Access to
the control room was controlled and operations personnel carried out
their assigned duties in an effective manner. Control room demeanor and
communications were appropriate.
Plant tours and perimeter walkdowns were conducted to verify equipment
operability, assess the general condition of plant equipment, and to
verify that radiological controls, fire protection controls, physical
protection controls, and equipment tagging procedures were properly
implemented.
Security Force Member Vacated Compensatory Post
The inspectors reviewed the licensee's investigation report concerning
an alarmed door to a vital area that was left unattended with the alarm
function non-operational.
The incident occurred on April 22, 1993.
This incident will be followup by a regional security inspector.
Radiation Monitor Deficiencies
IR 93-08 discussed a 10 CFR Part 21 Report concerning a deficiency in
the electrical design of the process and area radiation monitors
utilized at the site. Specifically, an internal 5 volt power supply
could fail -without -the--failure-being annunciated. -However, the loss of
this power supply would also result in the loss of the digital readout
and thus was readily detectable. The vendor, Nuclear Research
Corporation, designed a circuit change to correct the deficiency. The
change involved installation of a jumper on the printed circuit boards
associated with the ratemeters. By April 21, all 22 radiation monitors
had been modified. The inspectors witnessed portions of the tests
performed to return the radiation monitors to service. The modified
units appeared to function correctly.
3
A EDG Partial Start During Barring Evolution
IR 93-07 described an events in which during barring over of the A EDG,
the engine speed began to rapidly increase and the engine was shutdown
by the operator. The barring evolution after each EDG run was a vendor
recommendation. The barring clears lube oil from the top of the bottom
pistons, i.e., reduces the likelihood of an exhaust manifold fire during
a subsequent start.
During this report period, additional testing was performed to recreate
the events and determine probable cause. The licensee utilized a
Woodward governor vendor representative to assist in this effort. The
testing verified that the governor will move the fuel racks partially
open and then if the operator does not intervene, the fuel racks will
subsequently close. Thus, operator action was not required to shut the
engine down before it reached full speed. The inspectors witnessed
performance of some of the tests and reviewed the operation of the
governor with the vendor representative. From a schematic which shows
the hydraulic operation of the governor, the observed phenomena was not
explainable. No diagram was available which showed the actual flow
paths and design of the internal components and clearances. The vendor
representative indicated that the phenomena was not unheard of and that
adjustment to shutdown solenoid may correct this condition. Both the
vendor representative and the cognizant engineer continued to support
the licensee's position that thle phenomena did not adversely affect the
function of the governor, i.e., the A EDG. The inspectors agreed with
their assessment. Engineering was evaluating replacing the governor
with one from stock and shipping the governor to the vendor for bench
testing. The inspectors will continue to monitor the licensee's efforts
in this area.
Failure Of Capacitor C-8 in A EDG Voltage Regulator Circuit
On April 26, 1993, the inspectors observed that capacitor C8 on the EDG
A voltage regulator circuit board was damaged. At the time of this
observation, the EDG was inoperable as a result of ongoing maintenance
and testing. A subsequent inspection by the licensee confirmed the
inspectors observation and the capacitor was replaced later that day.
To ensure no collateral damage to other components in the voltage
regulator circuit, the licensee also performed a satisfactory resistance
check of diode CR8 in the voltage regulator circuit on April 28, 1993.
The EDG was returned to service on April 29, 1993, following
satisfactory operation of the voltage regulator at an EDG loading of
2500 KW during performance of OST-401.
Capacitor C8 was wired across a rectifier bridge in the automatic
voltage regulator circuit for the EDG. After analyzing the regulator
circuit and discussions with the voltage regulator vendor, the licensee
stated that the damaged capacitor acted as a filter for the output of
the rectifier bridge. The licensee also stated their conclusion that
failure of the capacitor would not adversely impact the ability of the
regulator circuit to perform its function. Additionally, the licensee
4
concluded that if collateral damage occurred to the circuit with the
failure of C8, the most likely component to be degraded would be diode
CR8.
The inspectors independently reviewed the circuit schematic and, after
discussions with NRC Region II staff, concurred with the licensee's
conclusion regarding the function of the capacitor. The inspectors also
concurred with the licensees conclusions regarding diode CR8 as the most
likely component for collateral damage. The inspectors witnessed the
replacement of the capacitor per WR/JO-AEWW1 and the restoration of EDG
A to service in accordance with OST-401. The inspectors also reviewed
the results of the satisfactory electrical checks on diode CR8. The
inspectors have no further question on this issue at this time.
B Fuel Oil Transfer Pump Frequent Cycling
At 9:36 p.m. on April 27, 1993, during an operability run of B EDG in
accordance with OST-409, operators noted that B fuel oil transfer pump
was cycling frequently. The pump was observed to run for 30 seconds out
of every two minutes. At the time of this observation, A EDG was
inoperable as a result of ongoing repair/testing activities associated
with failed capacitor C8 and the barring anomaly discussed in the
preceding paragraphs. An operability determination was initiated for
the B fuel transfer pump at 10:43 p.m. on April 27, 1993.
At 4:59 p.m.
on April 29, 1993, following the restoration of EDG A to service, B EDG
was declared inoperable and TS 3.7.2.d LCO was entered to allow repairs
to the day tank level circuitry. The LCO allowed operation to continue
for seven days before placing the reactor in hot shutdown. Following
replacement of the day tank low level switch, B EDG was returned to
service at 10:30 p.m. on April 29, 1993, and the TS LCO was exited. The
operability determination was completed at 10:18 p.m. on April 29, 1993,
and concluded that the fuel oil transfer pump and its associated
starting circuitry components would continue to operate satisfactorily
at the observed cycling frequency.
The inspectors independently reviewed the operability determination and
the completed work request and have no further question at this time.
No violations or deviations 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 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
5
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/reviewed portions of the following test activities:
Operating Procedure Diesel Generators A and B
(Section 8.6: Barring Over Diesel Generator A)
OST-051
Reactor Coolant System Leakage Evaluation
OST-352
Containment Spray System Component Test
OST-409
Emergency Diesels (Rapid Speed Start)
OST-401
Emergency Diesels (Slow Speed Start)
No violations or deviations were identified. Based on the information
obtained during the inspection, the area/program was adequately
implemented.
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 and approved procedures. The inspectors determined
that these activities did not violate LCOs and that required redundant
components were operable. The inspectors verified that required
administrative, material, testing, radiological, and fire prevention
controls were adhered to. In particular, the inspectors
observed/reviewed the following maintenance activities:
WR/JO 93-AEWW1
Replace Capacitor C-8 On A EDG
Voltage Regulator Circuit Card
WR/JO 93-ADFN1
Replace Mechanical Seal On A EDG
Standby Coolant Pump
WR/JO 93-AEMQ1
Troubleshoot A EDG Control Circuit
Inadequate Post-Maintenance Testing Of AMSAC
IR 93-07 discussed the March 31, 1993, failure of both AMSAC channels.
Due to the unavailability of spare parts, only the A channel was
repaired-and returned to service on April 10.--The repair was
accomplished by replacement of the A channel microprocessor.
Satisfactorily performance of the new microprocessor was considered to
be demonstrated by performance of SP-1198, AMSAC System Test (At Power).
On April 20, the inspectors reviewed SP-1198 and determined that it did
not functionally test the entire A channel.
During subsequent
discussions with engineering personnel, the inspectors were informed
that the circuitry not tested by SP-1198 was periodically, automatically
tested by AMSAC's self-test feature. The inspectors requested the
6
licensee provide documentation which would described the self-test
feature in sufficient detail such that one could verify that the entire
circuitry had been tested. While reviewing the applicable sections of
the AMSAC vendor manual, the licensee discovered that the self-test
feature of a channel was not performed when the other channel's
microprocessor is out of service. Since the B channel microprocessor
was not functional, portions of the A channel remained untested after it
was returned to service on April 10.
Essential elements of the A
channel that were not tested included the logic output contacts.
1198 was revised to perform a complete logic test of the A channel.
Satisfactorily performance of the revised SP-1198 was witnessed by the
inspectors on April 23.
Operability and testing of AMSAC was not addressed in TS. However, the
system was required to be installed by 10 CFR 50.62 and thus is
considered as important to safety. Failure to provide an adequate test
procedure for testing essential elements of the circuitry associated
with the replaced microprocessor constituted a failure to establish
procedures appropriate to the circumstances as required by 10 CFR 50
Appendix B Criterion V. This is identified as a VIO: Failure To
Establish Adequate Procedures To Verify Proper AMSAC Operation After
Microprocessor Replacement, 93-10-01.
As discussed above, the B channel was not returned to service at the
time the A channel was placed in service due to the unavailability of
another microprocessor. A microprocessor was obtained from the vendor,
Modicon Sealed Support Center, and the B channel was successfully tested
and returned to service on May 7. However, the vendor has no additional
microprocessors in stock and no longer manufactures this component.
Apparently, the AMSAC system installed at the site was installed in only
two other nuclear facilities. HBR was unable to obtain from either of
these two sites a spare microprocessor that was compatible with the
system installed here. Hence, for future failures availability of parts
to repair AMSAC in a timely manner was a concern. At the end of the
report period, the licensee had initiated efforts to address this
concern. This item is identified as an IFI: Lack Of Spare Parts Could
Result In Prolonged Unavailability Of AMSAC, 93-10-02.
One violations was identified. Except as noted above, the area/program
was adequately implemented.
6. Event Followup (92701, 93702)
At 6:13 p.m.-on May-12,.-1993, an NOUE was declared,. in accordance with
EAL-2 flowchart criteria, for RCS leakage greater than 10 gpm. At 8:54
p.m. A charging pump and associated piping was isolated after it was
determined to be the source of the leakage. Utilizing OST-051, the RCS
leakage rate was confirmed to less than 10 gpm, i.e., had decreased to
0.0796 gpm. The NOUE was exited at 10:36 p.m..
The inspectors were notified of the event via the licensee's beeper
notification system. The inspectors reported to the site and from
7
approximately 6:45 p.m. to 11:00 p.m. witnessed the licensee's event
response. The inspection included observations both in the control
room, the OSC, and the TSC. In general, the response was deemed
satisfactory. The inspectors also verified that state, local, and NRC
notifications were performed in accordance with regulatory requirements.
The source of the leakage was determined to be the A charging pump
suction relief valve, CVC-2080. An inspection of the valve by the
licensee revealed that the setpoint adjusting bolt nut had loosened from
its snug position against the bonnet face. This allowed the adjusting
bolt to loosen and resulted in a lowering of the valve lift setpoint
from 75 psig to 10 psig. The valve was disassembled and inspected;
however, the cause of the loose adjusting bolt nut could not be
definitely determined. The root cause of the loose adjusting nut will
be addressed by the licensee as part of the ACR process. During the
inspection light scoring was observed on the valve disc. Following
replacement of the disc, nozzle, and spindle, the valve was
satisfactorily retested and reinstalled in the system.
An inspection of the suction relief valves for the B and C charging
pumps indicated that the adjusting bolt nuts for those valves were in
the correct position. A review of plant and industry experience by the
licensee, as well as, discussions with the valve vendor failed to reveal
any prior occurrences of this problem. Thus, the licensee considered
this an isolated event.
The inspectors witnessed a portion of the repair and testing efforts
associated with the A charging pump suction relief valve. Additionally,
the inspectors reviewed WR/JO 92-AKWG1, under which the last maintenance
was performed on CVC-2080 on October 9, 1992.
No abnormalities were
identified in that WR/JO which could have resulted in the loose
adjusting nut.
On May 13, a licensee review identified that during the event, key
personnel had exceeded the working hour guidelines of TS 6.2.3.b.
without approval from the Plant General Manager or his designee.
Specifically, the TS guideline states that "An individual should not be
permitted to work more than ...
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period ...
excluding shift turnover time. Any deviation from the above guidelines
shall be authorized by the Plant General Manager or his designee ...
in
accordance with established procedures..." PLP-015, Program For Nuclear
Power Plant Staff Working Hours, implements this requirement. ACR 93
084 was initiated to address this TS violation. 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. Thus, this item is identified
as a NCV: Failure To Authorize Shift Work Hours In Excess Of Those
Specified In TS 6.2.3.b, 93-10-03.
One NCV was identified. Except as noted above, the area/program was
adequately implemented.
8
7. Loss Of Decay Heat Removal (TI 2515/103)
By letter dated February 1, 1989, the licensee described commitments to
address the six programmed enhancement recommendations identified in GL 88-17. The inspectors verified that M-1011, Instrumentation For Midloop
Operation, installed instrumentation for midloop operation as committed
in their February 1, 1989 letter. The inspectors also verified that
requirements for equipment availability and key parameter monitoring
were contained in GP-008, Draining The Reactor Coolant System, and OMM
030, Control Of CV Penetrations During Midloop Operation as required.
Current procedures did not allow perturbations of the RCS and thus
additional procedures were not necessary to address this area. The
inspectors also reviewed the February 1, 1989 response against the six
items discussed in GL 88-17. Based upon the inspection activities, the
inspectors concluded that licensee met the intent of GL 88-17 items 1
through 4 and 6 by a combination of procedure revisions and new hardware
installations. In response to item 5, the licensee determined that no
TS changes were required. This temporary instruction is considered
closed.
8. Followup (92701, 92702)
(Closed) VIO 92-11-01, Failure To Implement FP-005 Resulted In Alert
Declaration. The inspectors confirmed that FP-005, Hot Work Permit, was
revised to help ensure that all actions required to be performed prior
to work authorization are completed. Training records were reviewed to
confirm that designated plant personnel were trained on the procedure
revision. In addition, the inspectors verified that ACRs92-284, 285
and 291 were initiated to review other key plant work processes, i.e.,
RWPs, confined space and equipment clearance programs. The reviews
required by these ACRs have been completed. The inspectors noted that
action items were identified to address the ACR findings; however, the
inspectors did not review the adequacy of the proposed corrective
actions for the these findings. This item is considered closed.
(Closed) VIO 92-11-03, Failure To Implement Appropriate Instructions
During SW 374 and 376 Valve Maintenance. The inspectors verified that
the current revisions of MMM-001, Maintenance Administration Program,
and MMM-003, Maintenance Work Requests, contained the information
specified in the Reply To A Notice Of Violation, dated July 1, 1992.
Specifically, MMM-001 step 5.5.14 required Technical Support perform a
seismic review prior to removal of a safety related component that
results in loose piping or anchorage. MMM-003 Attachment 6.5, Work
Request-Planning-Checklist, referenced MMM-001-step 5.5.14 if a safety
related component is to be removed. These actions should preclude
recurrence of this violation. This item is considered closed.
(Closed) VIO 92-11-05, Failure To Translate RHR System Design Basis Into
M-1087. The inspectors verified that the following corrective actions
were adequately implemented as committed in the Reply To A Notice Of
Violation, dated July 1, 1992. RHR System DBD and SD-003, Residual Heat
Removal, were revised. A memorandum was issued to engineering personnel
9
requiring that when a modification is released to the plant for review,
a marked up copy of each affected plant procedure accompany the
transmittal.
NED procedure 3.3, Design Verification/Technical Review,
was established to institute a formal qualification program for
engineering personnel. This item is considered closed.
(Closed) VIO 92-16-03, CM-508 Was Not Adequately Established In That
Steps Provided For EDG Fuel Filter Assembly Were Out Of Sequence And
Failure To Adequately Establish Procedure CM-303 For EQ Splices. The
inspectors verified that CM-303 and CM-508 were revised as necessary to
provide adequate instructions for their respective activity. MI-506-0,
Maintenance Procedures Program, was implemented, as committed in the
Reply To A Notice Of Violation, dated August 20, 1992, to require
validation of procedure revisions. In addition, MI-506-0 addressed
tracking and trending of the validation process to determine the
effectiveness of the program. This item is consider closed.
(Closed) VIO 92-16-04, Instructions In M-1128 Were Not Appropriate To
The Circumstances In That The Modification Created An Unmonitored
Release Pathway. The inspectors reviewed the Reply To A Notice Of
Violation dated August 20, 1992. The inspectors verified that the
procedure were revised or in the process of being revised as committed
in the reply. Specifically, the inspectors verified that Attachment
6.3, Review Assignment Criteria, of AP-22, Document Change Procedure,
revision 11, dated March 27, 1993, required evaluation for unmonitored
release pathways. The inspectors also verified that a check sheet had
been issued for interim use until a similar change could be implemented
into the Nuclear Plant Modification Program manual.
These actions
should be sufficient to preclude recurrence of this event. This item is
considered closed.
No violations or deviations were identified. Based on the information
obtained during the inspection, the area/program was adequately
implemented.
9. Exit Interview (71701)
The inspection scope and findings were summarized on May 24, 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
93-10-01
VIO - Failure To Establish Adequate Procedures
To Verify Proper AMSAC Operation After
Microprocessor Replacement (paragraph 5).
10
93-10-02
IFI - Lack Of Spare Parts Could Result In
Prolonged Unavailability Of AMSAC (paragraph 5).
The following NCV was identified and reviewed during this inspection
period.
Item Number
Description/Reference Paragraph
93-10-03
Failure To Authorize Shift Work Hours In Excess
Of Those Specified In TS 6.2.3.b (paragraph 6).
10.
List of Acronyms and Initialisms
a.m.
Ante Meridiem
ACR
Adverse Condition Report
ATWS Mitigation System Actuation Circuitry
Administrative Procedure
CFR
Code of Federal Regulations
Corrective Maintenance
CV
Containment Vessel
CVC
Chemical & Volume Control
Design Basis Documentation
Emergency Action Level
Environmental Qualification
Fire Protection
gpm
Gallons Per Minute
GL
Generic Letter
General Procedure
HBR
H. B. Robinson
IFI
Inspector Followup Item
IR
Inspection Report
KW
Kilowatt
LCO
Limiting Condition for Operation
M
Modification
MI
Maintenance Instruction
MMM
Maintenance Management Manual
Non-cited Violation
NED
Nuclear Engineering Department
Notice of Unusual Event
NRC
Nuclear Regulatory Commission
OMM
Operations Management Manual
OP
Operations Procedure
Operations Support Center
OST
Operations Surveillance Test
p.m.
Post Meridiem
PLP
Plant Program
Psig
Pounds per square inch - gage
Radiation Work Permit
System Description
11
Special Procedure
TI
Temporary Instruction
TS
Technical Specification
Violation
WR/JO
Work Request/Job Order