ML14178A509
| ML14178A509 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/20/1994 |
| From: | Christensen H, Ogle C, William Orders NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A506 | List: |
| References | |
| 50-261-94-12, NUDOCS 9406060199 | |
| Download: ML14178A509 (10) | |
See also: IR 05000261/1994012
Text
Ekgt REGol
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report No.:
50-261/94-12
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: March 27 - April 23, 1994
Lead Inspector:
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W. T. Orders, Seni
Resident Inspector
Da e Signed
Other Inspector:
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C. R. Ogle, Residenf Inspector
Date Signed
Approved by:
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H. 0. Christensen, Chief
ate igned
Reactor 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,
engineered safety feature system walkdown, plant safety review committee
activities, and followup.
Results:
One violation was identified involving the licensee's failure to take adequate
corrective actions pertaining to a repeated occurrence of erroneously blocking
a fire door open (paragraph 3).
One unresolved item, was identified involving the licensee's interpretation of
closed systems outside containment (paragraph 7).
9406060199 940520
ADOCK 05000261
0
REPORT DETAILS
1.
Persons Contacted
- R. Barnett, Manager, Projects Management
- J. Benjamin, Manager, Project.Controls
S. Billings, Technical Aide, Regulatory Compliance
A. Carley, Manager, Communications
- B. Clark, Manager, Maintenance
- D. Crook, Senior Specialist, Regulatory Compliance
J. Eaddy, Manager, Environmental and Radiation Support
- D. Gudger, Specialist, Regulatory Affairs
S. Farmer, Manager, Engineering Programs, Technical Support
- J Harrison, Manager,
&RC Technical Support
B. Harward, Manager, Engineering Site Support, Nuclear Engineering
Department
- S. Hinnant, Vice President, Robinson Nuclear Project
- K. Jury, Manager, Licensing, Regulatory Programs
J. Kozyra, Licensing/Regulatory Programs
- R. Krich, Manager, Regulatory Affairs
A. McCauley, Manager, Electrical Systems, Technical Support
R. Moore, Acting Operations Manager
- P. Musser, Manager Engineering Assessment - Nuclear Assessment
Department
- M. Pearson, Plant General Manager
M. Scott, Manager, Reactor Systems, Technical Support
E. Shoemaker, Manager, Mechanical Systems, Technical Support
- D. Taylor, Plant Controller
- L. Woods, Manager, Technical Support
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
- Attended exit interview
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Plant Status
The unit began the report period with the plant stable at 90% power in
preparation for reactor physics testing. The unit went critical at
7:42 p.m. on March 20, and was placed on line at 11:55 p.m. the
following day. Following successful completion of reactor physics
testing, reactor power was increased to 95% on the morning of March 29,
and ultimately to 100% power on the morning of April 2, 1994. The unit
operated at or near full power until 10:14 p.m. on the evening of
April 3, 1994, when a large E-H fluid leak forced the operators to
manually trip the unit. Details pertaining to the reactor trip are
delineated in paragraph 3 of this report. After necessary repairs to
the E-H system were completed, the unit was made critical at 2:09 a.m.
on the morning of April 5, 1994, was placed on line that afternoon, and
achieved 100% power at 1:10 a.m. on the morning of April 7. The unit
2
operated at power for the remainder of the report period, although
power level was limited on selected days as a result of weir discharge
temperature limitations to Lake Robinson.
3. Operational Safety Verification (71707)
a.
General
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.
The inspectors reviewed shift logs, Operation's records, data
sheets, instrument traces, and records of equipment malfunctions
to verify equipment operability and compliance with TS. 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 through work observations and discussions with
Operations staff members. 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 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.
b.
Reactor Trip Review
On April 3, 1994, the unit was operating at 100% power. At
10:14 p.m. that evening, a manual reactor trip was initiated when
the operators determined the inability to recover from a loss of
E-H oil to the main turbine control system. The event sequence
started at 10:05 p.m., when the control room received a Turbine E
H Fluid Lo-Lo Level annunciator alarm. An investigation into the
cause of this alarm was initiated since the E-H Fluid Lo Level
alarm, which is supposed to be received before a Lo-Lo alarm, had
not been not received. At 10:07 p.m., a large leak was identified
on the #1 turbine governor valve (GV-1). The operators began a
turbine load reduction at a rate of approximately 5% per minute in
preparation for isolating the E-H oil supply to the valve. At
3
10:11 p.m. the E-H Fluid Reservoir Lo Level alarm was received,
followed shortly thereafter by an E-H fluid pump lockout which
caused the E-H oil pumps to trip. At approximately the same time,
the E-H Fluid System Hi/Lo Pressure alarm was received, and the
turbine governor valves began to drift closed as a result of the
loss of E-H oil pressure. The governor valves closing caused a
rapid load reduction which would have caused a turbine trip (from
governor valve closure) and a resultant reactor trip. This rapid
load reduction also resulted in increasing Reactor Coolant System
temperature and pressure which also would have caused an automatic reactor trip on over-temperature delta-temperature. In
anticipation of an imminent automatic reactor trip, the operators
initiated a manual reactor trip at 10:14 p.m.
The reactor protection system functioned as expected, although two
rod bottom lights did not illuminate to indicate that the rods had
fully inserted into the core. Operator action was taken to borate
the reactor for the worth of the two rods. It was subsequently
determined that the rods had actually inserted fully, and that the
problem was one of erroneous indication.
The inspectors concluded that the transient was managed
appropriately.
At the end of this report period, the inspectors had not completed
their review of the circumstances pertaining to the initiation of
the E-H oil leak or the reason for the failure of the E-H Lo level
alarm to initiate. These items will be addressed in report 50
261/94-15.
c.
Missing Support Bolts on Valves RC-553 and RC-516
On April 8, 1994, during a routine tour of the mechanical
penetration area, the inspector observed discrepancies in the
support bolting arrangement for valves RC-553 and RC-516 which are
containment isolation valves and isolate the pressurizer relief
tank from the gas analyzer. The inspector noted that of the 4
bolts provided to affix RC-553 to its dual supports, two were
missing and a third was backed out approximately 3 turns. In
addition, RC-516 was missing one bolt.
(The licensee subsequently
determined that this bolt was not missing, but the head was
sheared off.) In response to this observation, the licensee
entered an operability determination at 11:00 a.m. on April 8,
1994. At 11:08 p.m. that evening, NED calculations were completed
which were intended to demonstrate that the valves were operable,
and adequately supported in their present condition.
On Sunday, April 10, 1994, the inspector reviewed NED calculation,
RNP-C/SPPT-1983, Qualification of Support for Valves RC-553 and
516.
Based on the methodology used in the calculation, the
inspector advised the shift supervisor of concerns for the
4
operability of the valves. These concerns centered on the
calculation crediting the bolt which was backed out on RC-553 in
the calculation of the tensile stresses experienced by the support
bolts. Following this conversation, the inspector was contacted
by the on-call Engineering Technical Support Manager. During the
ensuing conversation, the inspector was advised that this concern
though not documented in the original calculation, had been
considered.
On April 11, 1994, the inspector was presented, and reviewed with
the preparing engineer, Revision 1 to the calculation which among
other things addressed the lack of full thread engagement for the
bolt on RC-553 and its impact on the tensile capacity of the
bolts. This revision also indicated that the valves were
On April 12, 1994, the inspector questioned the valve weight used
in the original version and first revision to the calculation.
Specifically, the inspector was concerned that additional hardware
attached to the valve support had not been included in the loading
calculations. This included items such as air regulators and air
solenoid valves. The licensee determined that this supplemental
equipment represented an additional 12 pounds of weight per valve.
Revision 2 demonstrated that despite this increase from the
original valve/actuator weight of 72 pounds, the valves remained
operable. The inspector reviewed a non-verified version of this
revision, presented to them on April 18, 1994, and have no further
questions on the calculation.
The inspector noted that two revisions of the calculation were
required to address routine questions pertaining to the
calculation. The failure to develop an original, stand-alone
version of the calculation to support the initial operability
determination is considered a weakness.
The inspector also questioned the fact that the observed valve
regulator air pressure setpoints for RC-553 and RC-516 exceeded
the 20 psig maximums specified on the valve drawing. The licensee
later informed the inspectors that discussions with the valve
vendor revealed that while the observed pressure readings of
approximately 35 psig exceeded the recommended pressures, the
valves were not damaged. The licensee also determined that no
formal procedure exists by which to set the regulators. The
licensee generated an ACR to resolve this issue and address any
potential generic implications of this finding.
d.
Inoperable Fire Door
On April 14, 1994, the inspectors noted that the automatic fire
door to pipe alley was inoperable as a result of it being blocked
by a pedestal-mounted, portable sign. The inspectors discussed
5
this observation with the Fire Protection Shift Technician who
confirmed the inoperable status of the door. He also stated that
he was previously unaware of this condition. Therefore, the
paperwork and checks necessary for declaring the door inoperable
had not been instituted. The fire technician removed the sign so
that the door would operate.
The inspectors noted that this occurrence was similar to a fire
door blockage described in IR 92-34 and the subject of NCV 92
34-02. In that event, the same door was obstructed by an air hose
placed across the door's threshold in anticipation of maintenance
activities in pipe alley. The corrective actions identified for
this first event failed to prevent this occurrence.
10 CFR 50, Appendix B, Criterion XVI, requires that corrective
actions be taken to preclude repetition of conditions adverse to
quality. Contrary to the above, the licensee failed to take
adequate corrective action to a December 23, 1992, automatic door
blockage in that a similar event occurred on April 14, 1994. This
is identified as a Violation, VIO: 94-12-01, Failure To Implement
Adequate Corrective Actions Results In Inoperable Fire Door.
4.
Maintenance Observation (62703)
The inspectors observed safety-related maintenance activities on systems
and components to ascertain that these activities were conducted in
accordance with TS, approved procedures, and appropriate industry codes
and standards. The inspectors determined that these activities did not
violate LCOs and that required redundant components were operable. The
inspectors verified that required administrative, material, testing,
radiological, and fire prevention controls were adhered to.
Specific
maintenance functions observed/reviewed by the inspectors included, but
were not limited to, the activities detailed below:
WR/JO 94-AEIPI
Maintenance Support For Airlock Test
WR/JO 94-BX0141
Calibrate SIS Loop Flow Transmitter (FT-
940 Only)
WR/JO 93-APCQ1
Spent Fuel Pit Skimmer Pump Maintenance
No violations or deviations were identified. The above maintenance
activities were performed satisfactorily.
5.
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
- 1
by qualified personnel in accordance with an approved test procedure,
6
test instrumentation was properly calibrated, the tests were completed
at the required frequency, and that the tests conformed to TS
requirements. Upon test completion, the inspectors verified the
recorded test data was complete, accurate, and met TS requirements, test
discrepancies were properly documented and rectified, and that the
systems were properly returned to service. Specific surveillance
functions observed/reviewed by the inspectors included, but were not
limited to, the activities detailed below:
a.
Containment Personnel Airlock Leakage Test
On April 6, 1994, the inspectors witnessed performance of EST-010,
Containment Personnel Airlock Leakage Test. Overall, the conduct
of the test was satisfactory and in accordance with the procedure.
The calculation of the airlock leak rate was independently
verified by the inspectors and matched that calculated by the
licensee. The inspectors also verified that the allowable
penetration pressurization system leakage was subsequently updated
on the control room minimum equipment list based on the results of
this test.
The inspectors observed that maintenance personnel involved in the
test appropriately questioned a discrepancy between a valve
description in the procedure and the valve label.
This
discrepancy was resolved with a temporary procedure change prior
to the continuation of the test. The inspectors also noted that
one of the individuals involved in the test stepped across a step
off pad at the entrance to the air lock without obeying the
instructions on the pad to remove his rubber shoecovers. The
inspectors identified this discrepancy to the technician involved.
A survey performed by the HP personnel after this occurrence
revealed no spread in contamination as a result of this lapse.
Later the step off pad was moved to the security door at the
vestibule to the air lock. The licensee generated an ACR to
address this issue. The inspectors have no further questions on
this testing.
No violations or deviations were identified.
6.
Quality Check Program Review
The resident inspectors performed a review of the CP&L Quality Check
program to evaluate its effectiveness. The inspectors reviewed Nuclear
Generation Group Manual Procedure NGGM 302-16, Administration Of The
CP&L Quality Check Program, and Nuclear Assessment Department Procedure
QLCH-01, Quality Check Program, during the review.
The purpose of the program is to provide a means for CP&L and contract
employees, who are involved in nuclear work, to report nuclear safety
related and quality concerns in a confidential manner. The Program is
designed to also accept other types of employee concerns such as
7
intimidation, harassment, wrongdoing, industrial safety, management
practices, etc. when an employee does not wish to go through the normal
management channels.
The program is designed to provide feedback of results of investigations
and evaluations to the submitter who reported the concern, if known.
The program is also designed to ensure the confidentiality of the
submitter's identification to the extent possible when processing and
investigating concerns and to safeguard information being maintained in
the Quality Check files.
Procedure QLCH-01, Quality Check Program requires, in part, that the
Manager - Quality Check, in conjunction with each nuclear plant Vice
President/Manager, and or off-site support Vice President/Manager will
ensure that employees terminating or transferring out of the Nuclear
Generating Group exit through the Quality Check Program. The procedure
also states that employees whose work activity is related to design,
procurement, maintenance, or operation and whose work activity could
impact the safe and reliable operation of the nuclear plant should exit
through Quality Check and complete an Employee Exit Questionnaire. The
procedure states further that a Quality Check representative will
attempt to contact employees who do not exit through the Quality Check
process to determine if they have concerns. Alternatively, the Quality
Check representative will try to get the names and addresses of such
employees for the purpose of mailing No-show Letters and Exit Interview
forms to them.
To determine the effectiveness of the program, the inspectors selected a
one week period of time during refueling outage 14 and requested that
the licensee provide information pertaining to the disposition of the
personnel exit interviews required by the Quality Check Program.
A review of the data requested for the period spanning from
September 19 - 25, revealed that 33 personnel terminated work for CP&L
during that one week period. It should be noted that all appeared to be
contractors.
Of those 33 people, the data indicates that none received an actual
interview before their departure, 29 were mailed employee exit
questionnaires, two underwent a telephone exit interviews, and two were
missed completely.
Of the 29 who were mailed employee exit questionnaires, the licensee
received no response from 13, 45 percent of the number mailed. Three of
the 13 were sent to invalid addresses and were returned by the post
office. The inspectors determined from the review, that the licensee
does not have a process in place to determine if the personnel to whom a
questionnaire was mailed but from whom no response was received,
actually received the questionnaire.
8
In summary, of the 33 personnel who departed Robinson during the period
of September 19 - 25, 1993, none received a quality check interview
before departure, and only 18 of the 33 or 55 percent were contacted.
The inspectors concluded that based on the data received, the quality
check program, as implemented at Robinson, was ineffective to" ...ensure
that employees terminating or transferring out of the NGG exit through
the Quality Check Program".
7.
Licensee Action on Previous Findings (92701, 90702)
(Closed) IFI 90-14-01, Assure Operability Determination Procedure
Adequacy.
The inspectors reviewed the licensee's procedure to operability
determinations, OMM-039, Operability Determination. The inspectors
consider that OMM-039 is adequate and have no further questions on this
issue. This item is closed.
(Closed) URI 90-17-02, Review The Basis For Discrepancies In Containment
Isolation Valves Identified In The FSAR and The Appendix J Test Program.
This URI documented discrepancies in containment isolation valves
between the FSAR and TMM-005, 10 CFR 50, Appendix "J" Testing Program.
To address the containment system design and basis, the licensee
developed Generic Issue Document, GID 90-181, Reactor Containment
Isolation. The inspectors reviewed this document and based on a review
of selected sections noted that it agrees with TMM-005. The inspectors
were informed by the licensee that a revision to the FSAR to incorporate
the information in the containment study is in progress. The licensee
stated that this revision would be promulgated in May 1994. Based on
the accomplished and planned effort, this item is closed.
During the review of this study, the inspectors noted that credit was
taken for closed systems outside containment which are normally vented
to the RWST. The inspectors will review the licensee's basis for this
approach. Pending this review this item will be tracked as an
Unresolved Item, URI: 94-12-02, Basis For Closed Systems Outside
Containment.
(Closed) VIO 92-27-01, Failure To Adequately Establish GP-008 To
Preclude The Loss Of Decay Heat Removal During RCS Inventory Reduction.
Inspection Report 93-27 documents VIO 92-27-01 regarding significant
deficiencies associated with a RCS draindown on September 12, 1992. The
inspectors reviewed the latest revisions to GP-008, Draining the Reactor
Coolant System and PLP-037, Conduct of Infrequently Performed Tests or
Evolutions. Both procedures now contain sufficient guidance and
instructions to correct the causal factors associated with the draindown
event. This item is closed.
9
(Closed) LER 93-011, Potential For Uncontrolled Release Due To Equipment
Hatch Seal Leak During Refueling.
This item was addressed in Inspection Report 93-21 and was the subject
of VIO 93-21-04. Thus, further inspection of this item will be
performed in conjunction with the closeout inspection of that violation.
This item is closed.
8.
Exit Interview (71701)
The inspection scope and findings were summarized on April 29, 1994,
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
VIO: 94-12-01
Failure To Implement Adequate Corrective Actions
Results In Inoperable Fire Door.
URI: 94-12-02
Basis For Closed Systems Outside Containment.
9. List of Acronyms and Initialisms
ACR
Adverse Condition Report
E-H
Electro Hydraulic
Final Safety Analysis Report
IFI
Inspector Followup Item
FT
Flow Transmitter
LCO
Limiting Condition of Operation
Non-cited Violation
NED
Nuclear Engineering Department
NGG
Nuclear Generation Group
OMM
Operations Management Manual
RC
Safety Injection System
TMM
Technical Management Manual
TS
Technical Specification
Unresolved Item
WO/JO
Work Request/Job Order