ML20134A345
| ML20134A345 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/29/1985 |
| From: | Cantrell F, Hehl C, Hopkins P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20134A334 | List: |
| References | |
| 50-395-85-37, NUDOCS 8511040362 | |
| Download: ML20134A345 (12) | |
See also: IR 05000395/1985037
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET,N.W.
ATLANTA, GEORGI A 30323
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Report No.:
50-395/85-37
~
-Licensee:
South Carolina Electric and Gas Company
Columbia, SC 29218
Docket No.:
50-395
License No.: NPF-12
Faci 1ity Name:
V. C.--Summer
Inspection Conducted:
September 7 - 30, 1985
. Inspector:
b h b ,I
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/0/2 7/gg
C.W.Hehl()
Q
Date Signed
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l0/29) $5
Perry C. H6pkins
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Da'te S'igned
Approved by:
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F. S. Cantrell, Sectioli(Cfb pf'
Date Signed
Division of Reactor Proj(cts
SUMMARY
Scope: This routine, unannounced inspection entailed 385 inspector-hours onsite
in -the areas of plant tours; operatio_nal safety verifications; monthly surveil-
lance observations; monthly maintenance observations; a review of the licensee
program' for _. maintenance modifications and design changes; and a review of
operating events.
Results: -One violation was identified - with the unit at power, a feedwater
isolation valve was rendered inoperable for a period of approximately ten days.
851104036285Oh95
ADOCK
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REPORT DETAILS
1.
Persons Contacted'
Licensee Employees
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Bradham, Director, Nuclear Plant Operations
- K. Woodward, Manager, Operations
B. Williams, Supervisor of Operations
- M. Quinton, Manager, Maintenance
- M. Browne, Manager, Technical Support
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- B. Croley, Group Manager, Technical and Support Services
- A. Paglia, Manager, Nuclear Licensing
- H. Sefick,' Associate Manager, Station Security
- P'.LaCoe, Nuclear Licensing
.
- G~
Putt, Manager, Scheduling and Maintenance
.
- R. Campbell, Engineer, ISEG
- H. :Donnelly, Senior Licensing Engineer, Licensing Engineering
'*A. Koon, Associate Manager, Regulatory Compliance
- M. Fowlkes, Engineer, Regulatory Compliance
- S. Hunt, Quality Assurance
.
Other licensee employees contacted included engineers,
technicians,
operators, mechanics, security force members, and office personnel.
- Attended exit interview
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2.
Exit Interview
The inspection scope and findings was summarized on September 27, 1985, with
those persons indicated in paragraph I above. The inspector described the
areas. insp'ected and discussed the inspection findings.
One violation was -
identified:
Violation. 50-395/85-37-01: With -the unit at power, a feedwater
isolation valve was rendered inoperable by isolation of its instrument
air supply for a period of approximately ten days.
~
The licensee acknowledged the inspection findings and did not identify as
proprietary any of the materials provided to or reviewed by the inspector
during the inspection.
During the exit interview, the licensee did state that they were concerned
-with a Trip Report discussing the results and findings from a site visit on
July 22 and 23, 1985, by two NRR persons and one Region II inspector
conducting a review of limited aspects of the licensee's training program.
The licensee expressed a concern that this report had been placed in the
Public Document Room without the licensee being forwarded a copy.
The
inspectors acknowledged the licensee's comment.
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3.
Operational Safety Verification (71707, 71710)
The inspector observed control room operations, reviewed applicable logs and
conducted discussions with control room operators during the report period.
The inspector verified the operability of selected emergency sy stems ,
reviewed removal and restoration logs, and tagout records, and verified
proper return to service of affected components.
Tours of the control,
auxiliary, intermediate, diesel generation, service water and turbine
buildings were -conducted to observe plant equipment conditions including
potential fire hazards, fluid leaks. and excessive vibrations, and to verify
. that maintenance requests had been initiated for equipment in need of
' maintenance. The inspector, by observation and direct interview, verified
that the physical security plan was being implemented in accordance with the
Station Security plan.
'Within the areas inspected, no violations or deviations were identified.
4.
Surveillance Observation (61726)
During the inspection period, the inspector verified by observation / review
that selected surveillances of safety-related systems or components. was
conducted in accordance with adequate procedures, test instrumentation was
calibrated, limiting conditions for operation were met, removal and
restoration of the affected components were accomplished, test results met
requirements and were reviewed by personnel other than the individual
directing' the test, and that any test deficiencies identified during the
testing were properly reviewed and resolved by appropriate management
personnel.
Within the areas inspected, no violations or deviations were identified.
5.
Maintenance Observation (62703)
Station maintenance activities of selected safety-related systems and
components were observed / reviewed to ascertain that they were conducted in
accordance with regulato ry requirements.
The following items were
considered in this review: the limiting conditions for operations were met;
activities were accomplished using approved procedures; functional testing
and/or calibrations were performed prior to returning components or systems
to service; . quality control record were maintained; activities were
accomplished by qualified personnel; parts and materials used were properly
certified;- and radiological
controls were implemented as required.
Maintenance Work Requests were reviewed to determine status of outstanding
jobs to assure that priority was assigned to safety-related equipment which
might affect system performance.
Within the areas inspected, no violations or deviations were identified.
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6.
Maintenance / Modifications and Design Changes (37700)
The inspectors reviewed four design changes and modifications, not
previously approved by NRR, (along with four that had been approved by NRR)
to ascertain that the changes had been appropriately reviewed and approved
by the licensee in accordance with section 6 of the Technical Specifications
(TS), and 10CFR50.59.
Item
Area
Modification No.
Subject
1
Reactor Control
MRF 20208
Reactor Trip Switch-
gear
~2
MRF 20249
Overpressure Protection
System
Modification
3
Plant Electrical
MRF 20448
Diesel Generator (DG)
Cor. trol and Relay Panel
Relocation
4
Radwaste
MRF 20172
Mixer Added to Waste
Monitor Tank
5
Nuclear Blowdown
MRF 20571
Nuclear Blowdown Holdup
System
Tank Stabilizer
Supports
6
Protection System
MRF 20354-A
Design Review, Westing-
(startup)
house CAB Drawing
Review
7
Reactor Heat
MRF 20435
Reactor Heat Removal
Removal
Calibrations Set Point
8
Electrical
MRF 20560
Electrical Fire Carrier
Certification
During the review, the inspectors verified (1) that the design changes were
reviewed and approved in accordance with TS and established QA/QC controls,
(2) that design changes were controlled by established procedures, (3) that
the licensee conducted a review and evaluation of test results, that these
test results were within previously established acceptance criteria, that
any test deviations were resolved and necessary retesting was accomplished
as appropriate, (4) that operating procedure modifications were made and
approved in accordance with Technical Specifications (TS), and (5) that
as-built drawings were changed to reflect the modifications.
For
modifications, the inspectors observed (1) that change activities were
conducted in accordance with the appropriate specifications, drawings, and
other requirements, (2) that acceptance and startup testing of modifications
were conducted in accordance with technically adequate and approved
procedures, and (3) the implementation of appropriate controls
(e.g.,
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firewatch, portable fire fighting equipment, welding and cutting permit,
etc.).
Additionally, the inspectors reviewed the outstanding facility
change requests (FCR) and determined that an excessive backlog was not
developing.
The following Technical Services Procedures and Station Administration
Procedures were reviewed:
Station Administration Procedure SAP 133, Design Control / Implementation,
Revision 3, April.25,' 1985
Technical Services TS 128, Initiation, Evaluation, and Approval of Design /
Modification Requests, March 1, 1984
Technical Services TS 129, Design Development / Design Package, March 1, 1984
Technical Services TS 130, Design Analyses and Calculations, March 1, 1984
.
Technical Services TS 131, Design Verification, March 1, 1984
Technical Services TS 132, Technical Services Disposition of Modification
Change Notices, March 1, 1984
Technical Services TS 133, Preparation of Interim Drawings / Sketches, March
1, 1984
Technical Services TS 134, Temporary Bypass, Jumper, and Lifted Lead Review,
October 12, 1984
Technical Services TS 135, Station . Design Change Interface and Implemen-
tation, March 1, 1984
A review was conducted of the Engineering Project Printouts to include
technical _ services tracking system.
It appears that the licensee has established and scheduled training and
familiarization with current and past changes / modifications etc., affecting
plant conditions 'and training for appropriate licensee personnel prior to
these personnel assuming responsibilities in the plant.
Based upon interviews with licensee personnel in engineering, licensing,
regulatory compliance, maintenance . and operations, interfaces appear to be
functioning adequately.
Generally, engineering had a good handle on the
total. design / modification process.
Further, a review / sampling of records reflect adequate control at all
levels.
Within the areas inspected no-violations or deviations were identified.
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7.
On-site Followup of Operating Events (93702)
During . this inspection period, the inspectors reviewed operating events
which occurred at the facility. The focus of this review was to ascertain
the safety significance of the event; evaluate performance of safety systems
and actions taken by the licensee; confirm that proper notifications of the
event, if required, had been made in accordance with 10CFR50.72; and
evaluate the need for further or continued NRC response to the event.
a.
Loss of Instrument Air Event
On September 8,
1985, the licensee reported to the NRC that, at
4:40 p.m. on September 8,1985, the Turbine Driven Emergency Feedwater
Pump, (TDEFW) was determined to be rolling at approximately 200 - 250
revolutions per minute (RPM) with no TDEFW start signal present and the
steam inlet valve (XVG-2030) indicating closed. The main steam header
supply valves to TDEFW pumps A and B, from B and C steam generators
respectively, were closed and the TDEFW speed was reduced to zero. The
determination that the TOEFW pump was rolling was made by the control
operators following problems witl. the instrument air (IA) compressors
which resulted in a short term reduction in IA header pressure.
At 4:40 with the unit at 100 percent power "B"
instrument air
compressor tripped on an indicated high vibration, high temperature and
high lube oil temperature. Upon indication that the
"B" IA compressor
trips, the operator started "A"
IA compressor.
"A"
IA compressor
started, but apparently failed to supply air as the IA header pressure
continued to decrease. The operator then attempted to place a third,
,
supplemental
air compressor,
in service.
The supplemental air
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compressor started, but tripped almost immediately on high discharge
pressure.
At 4:47
"B"
air compressor auto started on low header
pressure, control was transferred by the operators local manual
control, and IA header pressure was restored to a near normal valve of
95 PSIG.
Normal IA header pressure is approximately 110 PSIG.
Subsequent to the restoration of "B" IA compressor, a mobile diesel-
driven air compressor was hooked up to the IA header, but would not
start due to a dead battery.
Following stabilization of IA header
pressure, the TDEFW pump was declared inoperable at 5:30 p.m.
on
September 8,1985 based on apparent excessive seat leakage through the
steam inlet valve PVG-2030 allowing the turbine to roll at low RPM.
The low RPM makes the turbine unreliable due to concerns with poor
bearing lubrication and pressurizing the Woodward governor, with the
potential for an overspeed trip following receipt of an automatic start
signal. During subsequent trouble shooting by operations to determine
the scope of the TDEFW pump problems, the pump was observed to roll
even with the-trip throttle valve closed.
Subsequent evaluation and trouble shooting determined that steam inlet
valve PVG-2030 had experienced steam cutting of the valve seat allowing
excessive seat leakage.
The valve seat was subsequently repaired and
the valve returned to service.
The observed phenomenon of the TDEFW
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pump rolling with the trip throttle valve closed was determined to have
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A and B, TDEFW pump turbine throttle valves drain valves open two turns
to allow draining of moisture collecting in the above and below seat
,
areas of this valve. The maintaining of these drain valves two turns
open effectively short circuited the steam around the trip throttle
valve. The licensee's valve alignment procedures have been changed to
require the XVT-2803 B below seat drain to be maintained fully closed,
with any moisture collecting in this area now being drained through the
turbine casing drains.
The TDEFW pump was successfully tested and
returned to service at 10:10 a.m. on September 11, 1985.
Although the above problems with the TDEFW pump were identified
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- following the September 8,1985 partial loss of instrument air, the
licensee has not been able to definitively determine whether the TDEFW
pump was rolling at low RPM prior to the event. The steam inlet valve,
PVG-2030, fails open on loss of IA, but the air system for this valve
has an accumulator to allow limited operation of the valve following a
- loss of IA. Since the as found testing of the leak tightness of this
' accumulator and~ associated piping was not performed prior to its being
disturbed during repair of valve PVG-2030, a determination of the
effect of the partial loss of IA on September 8,1985 cannot be made.
Post maintenance testing of the PVG-2030 valve included testing of the
valve's ability to function upon isolation of the IA supply; excessive
leakage in the'IA system was not identified during tnese tests. Review
of the performance of Surveillance Test Procedure (STP) 120.002, TDEFW
Pump Test, a monthly STP requiring the running of the TDEFW pump and
includes operator verification that the TOEFW turbine rolls to a stop
following closure of valve PVG-2030, indicates that on August 26, 1985 *
the turbine rolled to a stop following closure of PVG-2030. Aside from
the performance of STP 120.002, routine periodic determination that the
,
TDEFW pump is not rolling, is not performed.
With' regard to the TDEFW pump reliability concerns associated with the
turbine rolling at-low RPM, prior to declaring the TDEFW operable on
September 11,1985, -the licensee's engineering organization evaluated
the bearing' lubrication concern and determined that no adverse affect
on system reliability was expected. The licensee engineering organi-
zation was pursuing but had ; not completed their evaluation of the
potential for a turbine trip on overspeed occurring as a result of the
TDEFW pump rolling at low RPM supplying oil to the governor speed
setting cylinder.
The potential reliability concern associated the
governor's ability to control turbine speed during a start when
previous shaft motion was present (as was the case on September 8,
1985) is. potentially generic concern identified following a December
1984 event at Crystal River Unit 3 and discussed in IE Report
50-302/85-04.
As identified above, 'following the initial trip of the running IA
compressor
"B"
on September 8,
1985, apparent multiple sequential
failures prevented the installed standby and the supplemental air
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compressor from performing their intended function.
The plant
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instrument air systems, although not a safety related system, are of
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concern in their potential 1 affect on the safety related components
served. By design, the plant instrument ai.r systems are not required
,
to . function to mitigate the consequences of an accident in that all
.
. valves are either designed to fail to the required accident mitigation
position or are provided with air accumulators. A loss of instrument
air should not affect the ability of safety systems ' to perform their
safety functions.
Unfortunately, . instrument air systems that are not
properly maintained have been known to precipitate corrosion and
' fouling related failures in safety related systems.
Thus maintenance
and reliability related problems in IA systems must be examined' closely
to ascertain if they are indicative of a lack of adequate preventative
maintenance on the IA system which could precipitate safety system
failures. A review of the V.C. Summer program to assure quality of IA
determined that the IA systems were periodically monitored only by the
F
Health Physics organization to assure the air met breathing air
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requirements.
The amount of moisture (dewpoint) in the IA system is
not periodically monitored to assure proper air dryer ope' ration. A
review of the maintenance performed on the IA compressors following the
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September 8,
1985 event did not identify any failure mechanisms
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directly attributed to IA quality.
Despite the absence of a program
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for monitoring the quality of instrument _ air, no apparent degradation
of safety related components has resulted.
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b.
September 20, 1985 Reactor Trip
On September 20, 1985, the licensee notified the NRC that at 8:58'p.m.
a reactor trip from 100 percent power had occurred in response to a
turbine trip initiated by a trip of all main feedwater (MFW) pumps.
The MR/ pump trip initiated during the conducted of a special test to
determine condensate pump net positive suction head (NPSH) in
preparation for a planned modification to convert these pumps from
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variable speed to constant speed units.
Following the reactor trip,
the plant responded as expected with the exception that the MFW
isolation valve XVG-16118 failed to close in response to the expected
automatic
isolation on low reactor coolant average
temperature coincident with P-4 (reactor trip breakers open), and
intermediate range nuclear instrumentation channel (NI) 36 failed
downscale. The feedwater isolation valve and NI-36 were repaired and
,
tested and the unit was returned to power on September 22, 1985.
The condensate pump test in progress which resulted in the reactor trip
was conducted to determine the NPSH of the condensate pumps at design
pump speed and maximum pump speed by taking manual speed control of
each pump and increasing pump speed while monitoring pump parameters.
To avert a trip of the condensate pumps on high discharge pressure
during the testing, the pressure transmitter in the common condensate
pump discharge header was isolated.
The isolation valve for this
transmitter apparently allowed sufficient leakage to allow- the
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transmitter to sense the elevated header pressure expected during the
testing-and-1nitiate a trip of the condensate pumps.
Subsequent . evaluation of the above identified failure of N1-36
downscale, following the reactor trip, was determined to have resulted
from excessive detector compensating voltage. The compensating voltage
was adjusted,'the channel tested and was returned to an operable status
-prior to plant restart.
During plant startup following this reactor trip, a misaligned
instrument block equalizing valve for flow transmitter IFT 3571
rendered the emergency feedwater (EFW) flow control valve IFV 3546
incapable of automatically isolating EFW flow from the TDEFW pump.
This finding and a discussion of the failure of the main feedwater
valve XVG-1611 B to close as identified above are discussed in detail
below,
c.
Misaligned Equalizing Valve
On September 20, 1985, the licensee found emergency feedwater header
flow transmitter IFT 03571 inoperable because of an open equalizing
valve on the transmitter manifold. With this transmitter inoperable
-isolation of emergency feedwater to steam generator B on high. flow was
also inoperable.
The licensee identified the discrepancy and took immediate action to
rectify the situation.
The design analysis descriptions mitigate the
conse'quences of such an event. That the licensee's Emergency Procedure
further ' mitigates the situation, and by review of related documents
indicate that this is a highly isolated case'with minimum consequences.
There appears to be no programmatic deficiencies, relative to this
occurrence, therefore, no notice of violation will be issued.
d.
Failure of MFW Isolation Valve to Close
Following the reactor trip on September 20, 1985, MFW isolation valve
(FWIV) XVG-16118 failed to close in response to a feedwater isolation
signal on low reactor coolant average temperature coincident with
permissive reactor trip breakers open. Subsequent licensee evaluation
determined that the manually operated instrument air supply valve to
valve XVG-16118 had been isolated and the associated air receivers had
bled down to a pressure insufficient- to allow valve closure.
The
instrument air supply valve was determined to have been closed on
September 10, 1985, in conjunction with the disposition of a previously
identified problem in a portion of the valve hydraulic circuit not
required for FWIV 1611B to close in performance of its accident
mitigation function.
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The FWIV at V.C. Summer use hydraulic pressure to open and close the
valve. Instrument air pressure is necessary to position a 4-way valve
that directs the hydraulic pressure to the under or above valve
actuating piston area. The 4-way valve does not move and the FWIV
fails as is on loss of instrument air.
The valve operator has two
redundant air / hydraulic circuits labeled active and stand-by.
The
active side circuit alone receives signals from the Engineering Safety
Features (ESF) system to provide FWIV functions required for safe
shutdown and accident mitigation.
The inactive side receives signals
for waterhammer and steam generator baffle bolting protection as well
as a permissive for forward flushing (feedwaterline warming) opening.
~ Thus the accident mitigation functions are performed by the active side
alone.
This particular design' the valve receiving only a train "A"
,
ESF signal, is an accepted configuration in that the feedwater
regulating valves receive both train "A" and train "B" ESF signals and
the main feedwater pumps receive a train
"B" ESF (safety injection)
trip signal.
The use of non-safety related components (feedwater
regulating valves and the main feedwater pumps) for redundant accident
mitigation in a steam break senario has been accepted by the NRC as
discussed in NUREG 0138, page 1 - 11.
The previous problem with the FWIV noted above which resulted in the
isolation of IA to valve XVG-1611 B involved a defective 4-way valve in
the stand-by hydraulic circuit. Pending replacement of this defective
component at the next shutdown the licensee performed an engineering
evaluation in accordance with 10CFR50.59 and determined that since the
active side of the hydraulic circuit was unaffected by the defect, the
valve remained operable.
The 10CFR50.59 evaluation, attached to
Non-conformance Notice (NCN) 2032, determined that the FWIV would meet
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its functional requirements if the active side (train "A") hydraulic
accumulator is pressurized to at least 4800 psig.
Although not addressed in the disposition of the inactive side problem,
the defective 4-way valve caused the inactive hydraulic circuit to be
at a reduced pressure resulting in the air driven hydraulic pump
running constantly in an attempt to change this side.
To reduce
unnecessary wear on this pump, on September 10, 1985, a decision was
made to isolate IA air to the pump and establish a twice per shift
surveillance of the pressure in the hydraulic accumulator in the active
side to ensure pressure was maintained via the hydraulic accumulator
The IA supply isolation valve was subsequently closed.
Review of the IA isolation aspect of this event and the decision
process was indeterminate in that documentation was not available to be
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examined. The IA valve was not removed from service under the tag out
program. Its closure was not addressed on the disposition of NCN 2032.
Thus apparently no formal evaluation was performed or deemed necessary
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at the time. Concern 'over the bleed down of the air accumulators was
apparently not considered. It is noteworthy to mention that currently
there is no installed instrumentation to monitor air reservoir
pressure.
By design, a loss of IA (a non-safety system) should not preclude the
ability of this valve to perform its accident mitigation function. _ Air
reservoirs (accumulators) with associated check valves are provided to
maintain the required approximately 75 psig air pressure necessary to
position the 4-way valves to allow valve closure. With IA isolated,
either leakage through the IA reservoir check valves or other FWIV
components allowed the IA pressure to bleed down to a pressure that was
insufficient to close valve XVG-1611B.
No data is available to.
determine the actual cause of this air bleed down as the licensee
performed no leakage testing on valve XVG-1611B's air circuits prior to
replacement of the defective inactive side 4-way vahe.
Subsequent
" snooping" of the IA circuits on September 24, 1985, did not identify
any leakage.
Technical Specification' (TS) 3.7.1.6 requires that with the plant in
Modes 1, 2 or 3, each feedwater isolation valve be operable.
For a
component to be considered operable, TS definition 1.18 specifies that
it must be capable of performing its specified functions. One of the
safety functions of the FWIV is to close and isolate feedwater in
response to an ESF signal calling for feedwater isolation.
As identified above, during the period from about September 10,
1985, until September 20, 1985, FWIV XVG-1611B was rendered inoperable
fo.r performance of its safety functions by isolation of its instrument
air supply.
During this period, the unit was operated at 100's power
and when called upon to function to limit cooldown of the reactor
coolant system subsequent to the September 20, 1985 reactor trip, it
did not function.
This is a violation (50-395/85-37-01).
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On April 30, 1985, the NRC issued Information Notice (IEN) 85-35,
Failure of Air Check Valves to Seat".
IFN 85-35 discussed a
potentially significant problem concerning Parker-Hannifin check valves
supplied as part of the valve operators on Anchor / Darling Main Steam
Isolation Valves (MSIV) and FWIV.
Tnese check valves, which are
installed on the IA supply 'line to the air reservoirs, have been
demonstrated to fail to seat during a slow depressurization of the IA
supply line. As a result the MSIV or FWIV air resersoir would bleed
down rendering the valve incapable of being closed.
IEN 85-35,
Attachment 2,
identified V.C.
Summer Sta_ tion as a plant with FWIV
supplied by Anchor / Darling Company with these Parker-Hennifin air check
valves.
Inspection review of the licensee's evaluation and actions associated
with this IEN determined that the licensee's Nuclear Engineering Group
had determined that FWIV-1611A, B and C had been supplied with these
type check valve as identified in their August 21, 1985 letter to the
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plant and recommended that these check valves be replaced at the
- "ea rl iest opportunity".
This initial evaluation of IEN 85-35 by
Nuclear Engineering did not address or put forward any justification
for continued operation of the facility pending replacement of these
check valves. In light of the above bleed down of the air reservoirs
for valve XVG-1611B, and the lack of data to exclude failure of the
~ check valve to seat as a cause, the licensee on September 25, 1985
completed an engineering evaluation to justify continued operation
until the valves are replaced. NRC review of this evaluation was still
in progress.
Replacement of these valves is scheduled during the
October - November 1985 refueling :;utage. Completion of NRC review of
this justification and replacement of valves is an Inspector Followup
Item 50-395/85-37-02.
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