ML14176A850
| ML14176A850 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/05/1990 |
| From: | Dance H, Garner L, Jury K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14176A849 | List: |
| References | |
| 50-261-90-03, 50-261-90-3, GL-89-04, GL-89-4, NUDOCS 9004230509 | |
| Download: ML14176A850 (11) | |
See also: IR 05000261/1990003
Text
S"'tREG(
UNITED STATES
0 oNUCLEAR
REGULATORY COMMISSION
A aREGION
II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report No.:
50-261/90-03
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
Inspection Conducted: February 11 - March 10, 1990
Inspectors:
_
_
_
V_
_ /U_71
L.
Garner
Senior Res dint Inspector
Date Signed
K. . Jury, Resident I spector
Dat Signed
Approved by:-
-
/10
H. C. Dance, Section Chief
Date Signed
Division of Reactor Projects
SUMMARY
Scope:
This routine, announced inspection was conducted in the areas of operational
safety verification, surveillance observation, maintenance observation,
rule compliance, onsite followup of events and onsite followup of written
reports of nonroutine events.
Results:
Weaknesses were identified in the specification of adequate post-maintenance
functional testing.
The SI system DBD incorrectly identified the SI-890A and B valves as containment
isolation valves.
A PNSC approved TS interpretation involving containment integrity was later
determined to be invalid and was cancelled.
00 42:*30509 9)0405
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1
F'DC
REPORT DETAILS
1. Persons Contacted
C. Baucom, Senior Specialist, Regulatory Compliance
- D. Baur, Manager, Quality Assurance
C. Bethea, Manager, Training
- W. Biggs, Manager, Site Engineering Support
D. Crook, Senior Specialist, Regulatory Compliance
- J. Curley, Manager, Environmental and Radiation Control
- C. Dietz, Manager, Robinson Nuclear Project
- S. Edwards, Senior Engineer, Technical Support
R. Femal, Shift Foreman, Operations
- S. Griggs, Technical Aide, Regulatory Compliance
- E. Harris, Director, Onsite Nuclear Safety
- J. Kloosterman, Director, Regulatory Compliance
D. Knight, Shift Foreman, Operations
R. Moore, Shift Foreman, Operations
- R. Morgan, Plant General Manager
- P. Odom, Project Specialist, Maintenance
- M. Page, Manager, Technical Support
- S. Pruitt, Senior Specialist, Technical Support
D. Quick, Manager, Plant Support
D. Seagle, Shift Foreman, Operations
- J* Sheppard, Manager, Operations
R. Smith, Manager, Maintenance
R. Steele, Shift Foreman, Operations
H. Young, Director, Quality Assurance/Quality Control
Other licensee employees contacted included technicians, operators,
mechanics, security force members, and office personnel.
- Attended exit interview on March 14, 1990
Acronyms and initialisms used throughout this report are listed in the
last paragraph of the inspection report.
2. 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
2
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-process 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 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 continued to be informal yet
effective.
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.
No violations or deviations were identified.
3. Monthly 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 performed prior to test initiation, testing was accomplished by
qualified personnel in accordance with an approved test procedure, test
instrumentation was properly calibrated, the tests were completed at the
required frequency, and that the tests conformed to TS requirements.
Upon
test completion, the inspectors verified the recorded test data was
complete, accurate,
and met TS requirements, test discrepancies were
properly documented and rectified, and that the systems were properly
returned to service.
Specifically, the inspectors witnessed/reviewed
portions of the following test activities:
OST-051 (revision 11)
Reactor Coolant System Leakage Evaluation
OST-406 (revision 2)
TSC/EOF/PAP Diesel Generator
OST-910 (revision 11)
Shutdown Diesel Generator
MST-007 (revision 8)
Reactor Coolant Low-Temperature Overpressure
Protection System Test
MST-552 (revision 6)
Turbine Redundant Overspeed Trip System Testing
No violations or deviations were identified.
3
4. Monthly 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, testing, and radiological
controls were adhered to. In particular, the inspectors observed/reviewed
the following maintenance activities:
PIC-301 (revision 1)
Pressure Switches and Vacuum Switches
WR/JO 90-ABUF1
Repair of SI-890B Bonnet to Body Leak
WR/JO 90-ACKK1
Repair Leak on PI-151A, Charging Pump C Pressure
Switch
Inadequate Post-maintenance Functional Testing of SI-890B
On February 6, 1990, WR/JO 90-ABUF1 was initiated to replace the gasket on
SI-890B, the B CS pump discharge check valve.
Boric acid crystals had
been observed around one of the bonnet studs.
Replacement of the bonnet
to body gasket involved removal of the flapper assembly which is attached
to the bonnet. The flapper assembly was visually inspected for degradation
and freedom of movement prior to reassembly. The seating surface was also
inspected.
Unsatisfactory material conditions were not observed.
After
reassembly, the bonnet to body connection was satisfactorily leak checked.
The discharge piping was then drained and an air test was performed to
demonstrate that the disc would partially open.
The inspection and
testing described is consistent with relief request no. 8 as granted by
the H. B. Robinson IST SER dated July 23, 1985.
On February 14, 1990, the inspectors reviewed the completed work package
and determined that appropriate post-maintenance functional testing was
not performed.
Upon review of the function of the CS discharge check
valves, it appeared that under certain postulated accident scenarios one
of the check valves would be relied upon to prevent containment atmosphere
from entering the common suction associated with the other CS pump, the SI
pumps and RHR pumps.
The scenarios of concern involve a postulated rapid
containment pressurization which automatically initiates the CS system
concurrent with either a failure of one of the CS pumps to start or the
manual securing of a CS pump without closure of any system MOV. Paragraph
4.9.11.1.1 of the SI system DBD states that check valves SI-890A and B
function as outside containment isolation valves for CV penetrations P-44
and P-45.
As such, post-maintenance testing would have to include testing
as required by 10 CFR 50 Appendix J.
As discussed above, no functional
testing was performed to quantify the amount of leakage, if any, after
reassembly of the SI-890B valve. This was discussed with the acting plant
general manager on February 14, 1990.
4
Review of the safety significance of this item indicates that the
probability of an accident is not increased but the consequences of
certain accidents could be exacerbated.
Gross leakage (e.g., a check
valve sticking open) could potentially air bind-one or more of the above
listed pumps.
However, the inspection performed on SI-890B provides a
reasonable level of confidence that the check valve would not exhibit this
failure mode if
the above postulated scenarios were to occur.
The
licensee demonstrated per calculation RNP-C/MECH-1070,
dated March 5,
1990, that leakage rates of 2.23 scf/hr or less would not result in air
introduction into the suction header during the injection phase of a large
break LOCA. During the recirculation phase of an accident the RHR pump(s)
supply pressure to the CS
pump suction line.
This supply pressure
precludes any CV atmosphere leakage through the CS check valves.
As a
comparison, the licensee indicated that the check valves were purchased to
MSS SP-61,
1961,
Pressure Testing of Steel Valves, which allows only 10
cubic centimeters per hour of water or 0.1 scf of air per hour per inch of
diameter of nominal valve size (SI-890A and B are six inch valves). The
inspectors agreed that for the small leakage rates discussed above, there
would be no adverse impact on the function of any (excluding containment
integrity) ESF system. However, the testing and the inspections performed
were not sufficient to assure such small leakage rates.
The licensee
contended that the testing and inspections supported the specified leak
rate.
It is not unreasonable however, to assume a higher leakage rate
(e.g.,
5 scfm or more) since leak testing was not performed.
Higher
leakage rates could result in introduction of gas into the suction side of
a running CS pump.
This could degrade pump performance, i.e., reduced
flow rates.
Reduced CS flow with minimum ESF equipment operating could
put the plant in an unanalyzed condition.
At the end of the report period, the licensee was in the process of
developing a leakage test for these check valves and plans to test them no
later than the September 1990 refueling outage or before if maintenance
must be performed on the check valves.
The NRC has agreed that the
licensee's proposed plan is acceptable due to the small probability of
occurrence of the postulated scenarios prior to the outage.
Prior to the inspectors raising the concern about the adequacy of the
post-maintenance testing performed,
the licensee had identified that
SI-890A and B valves perform two safety functions.
One function is "to
close to protect the pumps from reverse flow".
This was documented in
Calc # 89-29, revision 0, which was performed December 1, 1989, reviewed
December 4, 1989,
design verified February 12,
1990,
and approved on
February 14,
1990.
During the exit on March 14,
1990, the inspectors
proposed a NOV for failure to perform adequate post maintenance testing as
5
required by 10 CFR 50 Appendix B Criterion V. Subsequent to the exit, the
licensee provided the following paragraph concerning the proposed
violation.
"For valve SI-890B, performance of the tests which comply with the
inservice testing program requirements of 10 CFR 50.55a(g) and Plant
Technical Specifications constitutes adequate post-maintenance
testing.
As allowed by 10 CFR 50.55a(g), relief from ASME XI for
this valve (Relief Request No.
8) was addressed by CP&L and NRC.
This Relief Request identified valve disassembly as an acceptable
method of demonstrating proper valve operability. Valve disassembly,
visual inspection, a check for freedom of movement,
and partial
stroke testing were performed on SI-890B following maintenance and
serve as the basis for demonstrating the ability of the valve to
perform its opening and closing functions."
After consultation with Region II management,
it
was determined that
similar concerns raised during maintenance team inspections had been
identified as weaknesses and not as violations.
Thus,
the proposed
violation was withdrawn by the inspectors. The following weaknesses were
identified as a result of consideration of the licensee's position and
decision making processes:
1) The relief request from ASME Section XI failed to provide a basis
for not performing a reverse flow test.
2) Understanding that meeting GL 89-04 guidance would result in reverse
flow testing, the licensee elected to comply with their approved ASME
Section XI program as documented in TMM-004,
In-service Inspection
Testing.
Since Robinson 2 was listed as a Table 1 plant (GL 89-04
Table 1, Plants With SERs To Be Issued In The Near Future) it was
deemed acceptable to await the SER issuance prior to implementing GL 89-04 for corrective maintenance.
3) The licensee has stated that disassembly and inspection constitutes
adequate testing.
Response to question 15 contained in "Minutes Of
The Public Meetings On Generic Letter 89-04", dated October 25, 1989,
stated that "disassembly and inspection of a check valve is not
considered a test."
4) There
appears to be an erronous plant perspective involving
post-maintenance testing, e.g., specify the ASME Section XI Program
and/or TS surveillance tests associated with the components versus
accessing the scope of maintenance performed, determining what design
function may be affected, and then identifying appropriate functional
testing to demonstrate that the component can perform its intended
functions.
6
The inspectors reviewed several documents which identified that the
containment isolation valves for P-44 and P-45 are the manual operated
double disc gate valves designated SI-891A and B respectively, not the
SI-890A and B check valves.
The licensee plans to correct the error in
Review of the previously performed ILRT Test
(April 1987)
revealed that the CS piping was vented upstream of the
SI-880A, B, C, and D valves, the motor-operated pump discharge valves.
Hence, a combination of the check valves SI-890A and B, and the SI-880A,
B, C, and D valves formed a containment boundary during the ILRT.
The
inspector questioned whether the NRC approved design utilizing SI-891A and
B as containment isolation valves meets the 10 CFR 50 Appendix A GDC.
This question is being reviewed within the NRC and is identified as an
IFI:
Review CS Header CV Penetration Isolation Configuration with GDC,
90-03-01.
No violations or deviations were identified.
5. ATWS Rule Compliance (2500/20)
In response to the requirements imposed by 10 CFR 50.62, the licensee
installed the Westinghouse designed AMSAC.
This system provides a means
to automatically trip the turbine and actuate AFW flow in the event of a
complete loss of feedwater transient.
Per Westinghouse analysis,
documented in WCAP 8330, this mitigating action prevents RCS
overpressurization and exceeding DNB limits.
The AMSAC system was
installed during the 1988-1989 refueling outage.
The inspectors walked down selected portions of the AMSAC system.
This
walkdown included, but was not limited to: the controlling unit, the
safety-related signal isolators, relays, portions of cabling, the RTGB
controls,
and associated annunciators.
Additionally, the inspectors
verified proper wiring configuration for the signal isolators and selected
output relays.
A review was conducted of: selected procurement
documentation, bills of materials, receipt inspection documentation, the
isolators specification and qualification reports,. the installing
modification package
(M-942
ATWS Mitigation)
and applicable safety
evaluations,
DCNs,
weld data reports (including verification of QC
inspection), the modification's acceptance test, and the systems design
basis document, DBDR-85-080/00-1.
The inspectors also verified that the
FSAR was updated and that selected operating procedures were revised to
incorporate ATWS mitigation and AMSAC operation.
Conditions were not identified which would render the system inoperable
and plant operators have an acceptable understanding of the system's
purpose and operation.
One discrepancy was identified in that revision 1
of the isolators'
vendor qualification report,
EIP-QR-002,
had not
received all required reviews.
This discrepancy was not significant as
the only change between revision 0 and revision 1 was an administrative
7
change specifying the specific series (SC993)
that the qualification
report covered.
These changes had no impact on actual isolator
qualification.
The AMSAC system does not compromise the safety features of the existing
safety-related protection system and the licensee's design as endorsed
through the SER was being properly implemented with no major exceptions.
It appeared that proper configuration and control of installed instrumen
tation was established and being maintained.
Through discussions with
operation's staff it was determined that system bypassing occurs during
related maintenance.
There is continuous indication of the bypass status
in the control room. The system as installed appeared to meet 10 CFR 50.62
rule requirements and applicable QA controls were evidently adequately
applied during the design, procurement, installation, and testing.
As a
result of this inspection, this module is considered closed.
No violations or deviations were identified.
6. Onsite Followup of Events (93702)
On February 26,
1990,
numerous alarms were received on the LPMS system.
From 0047 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br /> to 1115 hours0.0129 days <br />0.31 hours <br />0.00184 weeks <br />4.242575e-4 months <br />, approximately 65 "events" or impacts, were
recorded on two reactor vessel head monitoring channels.
The licensee
originally believed the events to indicate a separated control rod guide
tube flexure; however, after analysis by Westinghouse, it was determined
that loose part impacts had not occurred.
The alarms were coming from
channel electrical noise.
The cause of the electrical noise was not
determined and subsequent to the 1115 hour0.0129 days <br />0.31 hours <br />0.00184 weeks <br />4.242575e-4 months <br /> event, no further alarms were
received.
The electrical noise was considered to be an anomaly and
according to Westinghouse, frequently occurrs at other units.
No violations or deviation were identified.
7. Onsite Followup of Written Reports of Nonroutine Events (92700)
(Closed)
LER 87-03 and LER 87-07 EQ Cable Splice Deficiencies.
The
specific items addressed in the reports were corrected.
Corrective
actions to prevent recurrence were not effective as documented by the
findings of Inspection Reports 87-10 and 87-19.
A subsequent Notice of
Violation and Proposed Imposition of Civil Penalty was issued on June 16,
1988, in this area.
See Inspection Report 89-26 for closeout of the
violation associated with cable splice deficiencies.
(Closed) LER 87-28 Diesel Generator B Air Start Failure While Diesel
Generator A Inoperable.
The failure to start was attributed to the air
start solenoids and/or check valves malfunctioning. The licensee replaced
the A and B EDG air start valves and check valves (WR 87-AQUK1, 88-ADZB1,
88-ADZW1 and 88-AEBQ1)
as committed in the LER.
However, a licensee
8
review of the LER in July 1989 discovered that the air start valves were
not in a regularly scheduled PM program as stated in the August 9, 1988,
LER supplement.
The inspectors verified that PM-406, revision 0, EDG Air
Start Solenoid Valve Inspection, was issued November 30, 1989, to correct
this deficiency.
(Closed) LER 87-27 Inoperability of Redundant Equipment Due to Inadvertent
Loss of Motor Control Center 6.
The event was attributed to the
accidental actuation of the MCC-6 feeder breaker trip button while
removing a protective cover.
The cover had been installed to preclude
accidental actuation of the trip button.
However, poor design did not
allow for easy removal of the cover.
The inspectors verified that a new
type cover has been installed over the trip buttons on both MCC-5 and 6
feeder breakers.
This corrective action should prevent recurrence of the
event.
(Closed) LER 90-04 Breach of Containment Integrity Due To Failure of the
Personnel Airlock Door. The inspector reviewed the licensee's proposed
corrective actions to periodically check airlock components. If properly
implemented, these actions should be sufficient to preclude recurrence.
The inspectors attended the PNSC on February 2, 1990, which approved TS
interpretation 90-001 involving the TS phrase "properly closed and sealed"
as applied to the airlock door. After the meeting, the inspectors voiced
concern about the potential for violating TS if this interpretation would
be used under other circumstances.
Subsequent discussions with NRR and
Region II personnel revealed that the interpretation was invalid.
The
licensee was informed and the TS interpretation was cancelled.
As
discussed in the subject LER,
an instance occurred where the Licensee
relied upon the interpretation
instead of entering TS 3.0 as required.
Voluntary entry into in TS 3.0 is typically strongly discouraged by the
NRC.
However, entry into TS 3.0 in the instances discussed in the LER
were considered appropriate due to of the short duration (2 to 3 minutes)
each time the outer airlock door was opened for passage of personnel and
equipment to repair the inner airlock door.
A PNSC action item, 90-02,
dated February 2, 1990 has been issued to submit a TS revision for
incorporation of an LCO for an inoperable airlock door.
(Closed)
P2187-01,
Colt Industries D/G Indicator Valve Plug Thread
Deterioration.
The inspectors verified that the subject plugs were
replaced on A and B EDG per WR 87-AHPA1 (December 1988) and WR 87-AHNZ1
(January 1989) respectively.
The inspectors also verified that these
plugs are to be replaced every other refueling outage per step 7.6.2.18 of
Emergency Diesel Generator Inspection. - Number 3, revision 5.
This satisfactorily implements the vendor recommendations contained in
Colt Industries/CP&L letter dated April 30, 1987.
No violations or deviations were identified.
9
8. Exit Interview (30703)
The inspection scope and findings were summarized on March 14, 1990, 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. As discussed in the exit meeting, the licensee
provided a written statement concerning a proposed violation.
Excluding
this position which is provided in paragraph 4, dissenting comments were
not received from the licensee. Proprietary information is not contained
in this report.
Item Number
Description/Reference Paragraph
90-03-01
IFI -
Review CS Header CV Penetration Isolation
Configuration with GDC 9. List of Acronyms and Initialisms
ATWS Mitigation System Actuation Circuit
American Society of Mechanical Engineers
Anticipated Transient Without Scram
CFR
Code of Federal Regulations
Corrective Maintenance
Carolina Power & Light
CV
Containment Vessel
Design Basis Document
Design Basis Document Reconstitution
DCN
Design Change Notice
D/G
Diesel Generator
Departure from Nucleate Boiling
Emergency Operation Facility
Environmental Qualification
Engineered Safety Feature
Final Safety Analyis Report
GDC
General Design Criteria
GL
Generic Letter
IFI
Inspector Followup Item
Inservice Testing
LCO
Limiting Conditions for Operation
LER
Licensee Event Report
Loss of Coolant Accident
LPMS
Loose Parts Monitoring System
M
Modification
10
Motor Control Center
Motor Operated Valve
MSS
Manufacturer's Standardization Society
Maintenacne Surveillance Test
Nuclear Reactor Regulation
NRC
Nuclear Regulatory Commission
OST
Operations Surveillance Test
Personnel Access Portal
Process Instrument Calculation
Preventative Maintenance
PNSC
Plant Nuclear Safety Committee
Quality Assurance
Quality Control
Robinson Nuclear Project
Reactor Turbine Generator Board
scf
Standard Cubic Feet
Standard Cubic Feet Per Minute
Safety Evaluation Report
SI.
Safety Injection
TMM
Technical Support Management Manual
TS
Technical Specification
Westinghouse Corporate Atomic Power
W/R
Work Request
WR/JO
Work Request/Job Orde