ML18152A354
| ML18152A354 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/27/1995 |
| From: | Belisle G, Branch M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A355 | List: |
| References | |
| 50-280-95-05, 50-280-95-5, 50-281-95-05, 50-281-95-5, NUDOCS 9504110094 | |
| Download: ML18152A354 (17) | |
See also: IR 05000212/2003004
Text
Report Nos. :
- JNITED STATES
>JUCLEAR REGULATORY COMMISSiON
r1EGION II
101 MARIETTA STREET. N.W., SUITE 2900
ATLANTA. GEORGIA 30323-0199
50-280/95-05 and 50-281/95-05
Licensee:
Virginia Electric and Power Company
Innsbrook Technical Center
5000 Dominion Boulevard
Glen Allen, VA
23060
Docket Nos.:
50-280 and 50-281
License Nos.:
Facility Name:
Surry 1 and 2
Inspection Conducted:
February 12 through March 4, 1995
Lead Inspector:
i.':~r~~r k~dent Inspect
Inspectors:
L. W. Garner, Project Engineer
Approved by:
D. M. Kern, Resident Inspector
S. G. Tingen, Resident Inspector
£~ /
/1
'
'
.
- E7 llK
G. if2 Bel i s1aition Chief
Reactor Projects Section 2A
Division of Reactor Projects
SUMMARY
Scope:
3- )_f -yJ-:-
Date Signed
This routine resident inspection was conducted on site in the areas of plant
status, operational safety/refueling activities verification, maintenance
inspections, surveillance inspections, on-site engineering, plant support,
Licensee Event Report followup, and action on previous inspection items.
Inspections of backshift and weekend activities were conducted on
February 15, 28 and March 1, 1995 *
950411()094 950328
- DR
ADOCK 05000280
2
Results:
Operations:
The reactor vessel head lift was conducted in an efficient and professional
manner (paragraph 3.1.1).
Weaknesses were noted in the procedure used to establish refueling containment
integrity for fuel off-load.
The procedure specified an incorrect method for
establishing refueling containment integrity for some penetrations
(paragraph 3.1.2).
A non-cited violation was identified for failure to monitor the lo~d cell as
required by procedures when fuel was lowered into the spent fuel pool storage
location (paragraph 3.1.3).
Spent fuel pool parameters were properly monitored and maintained while the
of f-1 oaded Un it 2 fue 1 was stored in the spent fue 1 poo 1. ( paragraph 3 .1. 4).
Maintenance:
The scaffolding erected and equipment staged prior to the refueling outage did
not impact operation of Unit 2 (paragraph 4.1) .
Although damaged plastit chain barriers were observed, the oversight and
controls of switchyard activities were appropriate (paragraph 4.2).
Refueling calibration surveillances were performed within the required
interval (paragraph 5).
Plant Support:
During tours of the Unit 2 containment, good ALARA planning and training were
evident (paragraph 7.2).
r
I l
l
1.
Persons Contacted
Licensee Employees
REPORT DETAILS
- W. Benthall, Supervisor, Licensing
H. Blake, Jr., Superintendent of Nuclear Site Services
- R. Blount, Superintendent of Maintenance
- D. Christian, Station Manager
J. Costello, Station Coordinator, Emergency Preparedness
D. Erickson, Superintendent of Radiation Protection
B. Hayes, Supervisor, Quality Assurance
- D. Hayes, Supervisor of Administrative Services
C. Luffman, Superintendent, Security
- J. McCarthy, Assistant Station Manager
- A. Price, Assistant Station Manager
- S. Sarver, Superintendent of Operations
+R. Saunders, Vice President, Nuclear Operations
- K. Sloane, Superintendent of Outage and Planning
E. Smith, Site Quality Assurance Manager
- T. Sowers, Superintendent of Engineering
- J. Swientoniewski, Supervisor, Station Nuclear Safety
G. Woodzell, Nuclear Training
Other licensee employees contacted included plant managers and
supervisors, operators, engineers, technicians, mechanics, security
force members, and office personnel.
NRC Personnel
- M. Branch, Senior Resident Inspector*
D. Kern, Resident Inspector
- S. Tingen, Resident Inspector
- Attended Exit Interview
+Participated in Exit Interview By Telephone
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2.
Plant Status
Unit 1 operated at power during the entire inspection period.
Unit 2 remained shutdown for a planned RFO during the entire inspection
period .
L
3.
2
Operational Safety/Refueling Activities Verification (71707, 60710)
The inspectors conducted frequent tours of the control room to verify
proper staffing, operator attentiveness and adherence to approved
procedures.
The inspectors attended plant status meetings and reviewed
operator logs on a daily basis to verify operational safety and
compliance with TSs and to maintain overall facility operational
aware~ess.
Instrumentation and ECCS lineups were periodically reviewed
from control room indications to assess operability.
Frequent plant
tours were conducted to observe equipment status, fire protection
programs, radiological work practices, plant security programs and
housekeeping.
Deviation reports were reviewed to assure that potential
safety concerns were properly addressed and reported.
3.1
Unit 2 Refueling Activities
The licensee conducted refueling activities from February 12 to
March 4.
Refueling evolutions were performed by contractors under
the direct supervision of licensed personnel.
The inspectors
verified initial plant conditions, attended the pre-evolution
briefs, and observed refueling activities to determine whether
license conditions were*appropriately addressed .
3.1.1 Reactor Vessel Head Lift
On February 12, the reactor vessel head was removed and the
refueling cavity was filled with water in preparation for
fuel off-load. The inspectors observed that the pre-
evolution brief was detailed with communications, personnel
safety, and RP concerns being clearly emphasized.
A
licensed SRO was assigned to coordinate the head lift
evolution.
The inspectors toured the control room and
containment prior to the start of the head lift. Required
plant conditions (i.e., nuclear instrumentation, containment
integrity, RCS boron concentration, shutdown margin, etc.)
were properly established and maintained as required by
procedure 2-0P-FH-001, Refueling Operations, revision 2.
The inspectors inspected the refueling cavity area prior to
th~ head lift to determine whether VPAP-1302, Foreign
Material Exclusion Program, revision 6, was properly
implemented. A barrier was established around the refueling
area. All personnel and material which*entered the
refueling area were tracked using the refueling area
accountability log. Access to the area was kept to a
minimum.
The type and number of items brought into the
refueling area were properly packaged and handled.
The
inspectors considered that the FME controls were good .
Activitie.s inside containment were effectively coordinated
by the SRO in charge. Clear communications were established
I
3
between the main control room, the crane operator, and the
refueling floor.
The containment coordinator and RP
personnel directed all personnel not involved with the RV
head lift to leave containment.
RP technicians performed
continuous area dose rate surveys during the RV head lift
and effectively controlled access to minimize personnel
radiation exposure.
RP support inside containment was
comprehensive.
The RV head movement to the storage stand in
the containment basement level was effectively executed.
The inspector concluded that the RV head lift was conducted
in an efficient, professional manner.
3.1.2 Refueling Containment Integrity
The inspectors reviewed control room indications for the
containment isolation valves used to establish refueling
containment integrity required by TS 3.10.A.l.
Proce~ure
2-0PT-CT-210, Refueling Containment Integrity, revision 5,
was the controlling procedure.
The procedure contained
sheets which depicted the different penetrations and
provided instructions to ensure that the barriers were
acceptable.
The procedure provided for many contingencies
that might be encountered by the operators in establishing
integrity when other activities were in progress.
For the most part, refueling containment integrity was
established by tagging valves in the required position for
the entire period that integrity would be required.
However, there were some valves that were open during the
fuel movement evolution. The inspectors noted that
containment sump isolation valves 2-DA-TV-203A/B were open.
These two valves are designed to automatically close upon
receipt of a containment isolation signal, but the logic for
automatic closure of these valves was deenergized.
TS 2.10.A.l sta.tes in part, "For those penetrations which
provide a direct path from containment atmosphere to the
outside atmosphere, the automatic containment valves shall
be operable or the penetration shall be closed by a valve,
blind flange, or equivalent".
The inspectors questioned whether containment integrity was
properly established for refueling since the automatic
closure of the open isolation valves was not operable. The
SRO indicated that the valves remained operable since they
could be remotely shut from the control room.
The operators
also informed the inspectors that procedure 2-0PT-CT-210
allowed this specific configuration and that they had always
considered that integrity was acceptable with these
isolation valves open during refueling *
The inspectors did not agree with the licensee's
interpretation that valves capable of being remotely closed
,
4
constituted acceptable refueling containment integrity.: The
licensee's position was discussed with cognizant NRC staff
who indicated that remote manual closure capability did not
meet TS 2.10.A.1 for the small number of containment
penetrations that provide a direct path from the containment
atmosphere to the environment.
The licensee was informed
that their position was not acceptable.
2-0PT-CT-210 was
reviewed by the licensee and revised.
The procedure
revision included a list of penetrations where the TS would
apply and an acceptable isolation method for establishing
refueling containment integrity. These penetrations would
either be closed, water sealed, pressurized, or isolable by
an automatic valve actuated by a high radiation signal. The
inspectors noted that the containment sump penetration
valves that were open during the core off-load were
considered to be water sealed and therefore refueling
containment integrity had been established.
The previous 2-0PT-CT-210 instructions that implemented the
licensee's posftion that automatic containment valves
remained operable if they could be remotely shut from the
control room was identified as a weakness.
The inspectors
reviewed 2-0PT-CT-210, revision 6, prior to core on-load and
did not identify additional problems .
3.1.3 Fuel Off-Load
On February 13-15, the core was completely off-loaded using
controlling procedure 2-0P-FH-001.
The inspectors observed
off-load activities from the control room, the refueling
floor in containment, and the fuel building.
The inspectors
independently confirmed that TS requirements to begin fuel
movement were satisfied. Operators established the required
refueling cavity level band and monitored level using a cold
calibrated pressurizer level instrument.
This instrument
had been recalibrated to provide a control room alarm in the
event of decreasing refueling cavity level.
The main
control room was placed on emergency ventilation and direct
communications were established between the control room,
the refueling manipulator crane, and the fuel building.
The order of fuel movement was directed by the refueling
coordinator who tracked core status from the control room.
Fuel movements were specified in the order listed in the
refuel report. Communications were clear and fuel element
transfer times were consistent. The inspectors discussed
the refuel report with the refuel coordinator and determined
that fuel movements were being properly directed and
tracked *
A licensed SRO supervised fuel movements in the containment.
The inspectors discussed individual duties with contract
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personnel operating the fuel manipulator crane and the fuel
transfer system.
Contractor personnel performed the fuel
movement in accordance with approved station procedures and
were knowledgeable of their responsibilities. The crane
operator closely monitored load cell tension when
raising/lowering the fuel gripper tool and when moving fuel
assemblies.
The refueling SRO and contractor personnel
independently verified core locations and positive latching
of fuel assemblies.
The FME area coordinator performed
frequent inspections of the refueling cavity area to confirm
that the area remained clear of foreign material. The
inspectors checked a sample of the items located inside the
FME boundary, and confirmed that each was properly tracked
in the refueling area accountability log.
The inspectors
concluded that the fuel movement within containment was
performed safely under close SRO supervision.
When monitoring the refueling activities in the fuel
building, the inspectors noted that only one person was on
the crane's bridge.
The person operating the bridge crane
was guiding the fuel element into the storage location with
one hand while he operated the electric hoist with the other
hand.
He was watching the fuel element the entire time from
when it entered the storage location to when it appeared to
be on the bottom.
The inspectors noted that refueling
personnel were not monitoring the weight load cell that was
attached to the rigging between the hoist hook and the
refueling tool.
The inspectors questioned the corporate NFA
person who was present in the fuel building as to the
acceptability of the practice of not monitoring the load
cell and the number of required individuals needed to
perform the evolution.
The NFA person indicated that he was
not sure of the Surry requirements but that at North Anna,
because of equipment design, two people were required and
they were also required to monitor the load cell. The
inspectors immediately notified operations management as to
the conditions observed. Activities were stopped and an
additional person was assigned to monitor the load cell.
Deviation Report S-95-0372 was written and later that
evening refueling activities were halted again for the
installation of a larger load cell that could be easily
monitored.
The inspectors reviewed refueling procedure O-OP-4.8, Spent
Fuel Assembly Handling Tool, revision 6.
Precaution 4.7 and
Step 5.3.2 required that the load cell be continuously
monitored while the fuel element was being lowered into the
storage location to detect any binding.
TS 6.4.0 requires
that refueling procedures be followed.
The failure to
monitor the load cell in accordance with O-OP-4.8 was
identified as NCV 50-281/95-05-01, Failure to Monitor load
Cell. This NRC identified violation is not being cited
6
because criteria specified in Section VII.B of the NRC
Enforcement Policy were satisfied.
3.1.4 Spent Fuel Pool Parameters
The inspectors monitored fuel pool parameters while the
Unit 2 fuel was temporarily stored in the spent fuel pool.
The inspectors verified the following:
Spent fuel pool water temperature, level, and boron
concentration were being properly maintained.
At least one spent fuel pool cooling pump and heat
exchanger were operating and the other spent fuel pool
cooling pump and heat exchanger were operable.
Spent fuel pool temperature and level instrumentation
and accompanying control room alarms were calibrated.
The spent fuel pool makeup rate was approximately 500
gallons per day.
The inspectors concluded that the spent
fuel pool parameters were properly maintained.
3.1.5 Fuel On-load
3.2
On February 28 and March 1, the inspectors monitored fuel
movement from the spent fuel pool to the Unit 2 reactor
vessel.
The inspectors verified that procedures for fuel
movement were followed, containment integrity was
established, ventilation was in the refueling alignment, FME
in the fuel building and containment refueling areas was
properly maintained and an SRO was stationed in containment
during fuel movement.
The inspectors concluded that
refueling operations were conducted in accordance with TS
requirements and that command and control were good.*
Unit 2 CV Integrity Inspection
With the assistance of the cognizant system engineer, the
inspectors examined accessible portions of the Unit 2 CV
liner for degradation. Special attention was place on
observing the areas where the CV liner and floor meet for
signs of pitting, general corrosion or wastage of CV liner
material due to abrasion.
No areas warranting repair were
found.
The inspectors noted that the caulking material and
paint at the floor's surface to CV liner interface had
separated from the CV liner at some locations to form
crevices.
In places the crevices allowed the steel CV liner
to be exposed and were sites where potential corrosive
materials might accumulate.
The inspectors noted that the
CV liner behind the containment sump was generally not
assessable for inspection since normally installed trash
,
7
racks and screens preclude access.
The system engineer
indicated, that to his knowledge, inspections performed per
l/2-NPT-CT-101, Reactor Containment Building Integrated Leak
Rate Test (Type A Containment Testing}, had n~ver included
the area were the CV sump is located.
The inspectors noted
that on occasions these racks and screens are removed for
various activities, and thus, this area becomes accessible
for short periods of time in which inspections could be
performed.
These observations were discussed with plant
management.
Within the areas inspected, one NCV was identified.
4.
Maintenance Inspections (62703}
During the reporting period, the inspectors reviewed the following
maintenance activities to assure compliance with appropriate procedures.
4.1
4.2
Unit 2 Pre-outage Sc~ffolding and Equipment Staging
The inspectors walked down areas of Unit 2 prior to the shutdown
for the RFO in order to evaluate the impact of early scaffolding
and equipment staging.
The Unit 2 turbine building had a
significant amount of scaffolding installed in order to support
replacing the MFW heaters and main turbine maintenance.
Scaffolding and equipment staged in the remaining areas of Unit 2
were minimal.
The inspectors noted that barriers were installed
to protect sensitive equipment in the areas where scaffolding was
installed. The inspectors concluded that early scaffolding and
equipment staging did not impact safe operation of Unit 2."
Review of Switchyard Work
The inspectors toured the Surry switchyard during periods of
extensive switchyard work.
The inspectors noted that switchyard
management personnel were present and were sensitive to the
consequences of their actions. The inspectors noted that the
entrance gate was locked and appropriate signs were attached to
inform personnel of risks.
The work in progress was the
installation of a third SOOkv ring bus to complement existing
SOOkv buses 1 and 2.
The inspectors noted vehicles moving within
the yard and observed that personnel were self-monitoring their
activities. The inspectors did note, however, that many small
yellow plastic chains used as vehicle barriers were damaged and
were no longer effective. The Assistant Stafion Manager, who
accompanied the inspectors, also noted the condition of the chains
and informed the switchyard management of the problem.
The
inspectors considered that the plastic chains provided only a
visible and not a physical barrier to vehicle movement.
The
licensee was reevaluating the barriers and considering a change in
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material type.
The inspectors considered that the oversight and
controls of switchyard activities were appropriate.
Within the areas inspected, no violations or deviations were identified.
5.
Surveillance Inspections (61726}
During the reporting period, the inspectors reviewed surveillance
activities to assure compliance with the appropriate procedure and TS
requirements.
The inspectors reviewed the following TS refueling calibration
surveillances in order to verify that they were completed within the
required TS surveillance interval:
2-IPT-CC-CS-L-200B, RWST Level Loop L-200B Channel Calibration,
revision 1
2-IPT-SI-L-922, SI Accumulator Tank 2-SI-TK-lA Level Loop L-2-922
Channel Calibration, revision O
2-PT-26, RCS Pressure (P-2-403}, revision 4
2-PT-2.5, SG Level (L-2-474}, revision 2
TS 4.1 requires that these surveillances be performed on a refueling
cycle interval. The inspectors reviewed the performance copies of these
surveillances performed during the Unit 2 1993 RFO and verified that
they were completed.
The inspectors also verified that these
surveillances were being performed during the present RFO.
The
inspectors concluded that these surveillances were being performed in
accordance with the required TS surveillance interval.
Within the areas inspected, no violations or deviations were identified.
6.
On-Site Engineering (37551}
On February 23 and 24, the A and B RSSTs were deenergized to support
the switchyard bus 5 outage. This maintenance required that the
Unit 1 J* emergency bus be powered from the Unit 2 500Kv bus via the
station service transformers. During the time that this backfeed lineup
was in effect, voltage on the Unit 1 J emergency bus was higher than
normal.
The inspectors reviewed TS 3.16.B.4 which allows a unit to be
operated for up to seven days when a primary off-site power source is -
not available.
When a primary off-site power source is not available, a
dependable alternate source must be operable. While Unit 1 was in a
backfeed alignment, the action statement for TS 3.16.B.4 was entered.
The inspectors questioned if equipment operation had been evaluated with
the higher than normal voltages on the emergency bus.
The inspectors
reviewed Engineering Report NP-1912, Evaluate Station Voltage, dated
January 30, 1991.
The report stated that during normal electrical
9
lineups maximum emergency bus voltages are 509V and 4305V on the 480V
and 4Kv buses respectively. During the backfeed lineup, emergency bus
voltages could be as high as 554V and 4683Kv on the 480V and 4Kv buses,
respectively. This report concluded that cable and switchgear ratings
were typically 600V for 480V equipment and 5Kv for 4160V equipment and
that overvoltages conditions were therefore not a concern.
The report
also stated that higher than normal operating temperatures could occur
in motors when operated at increased voltages and the time on backfeed
should be minimized.
The inspectors concluded that operation of the
unit in a backfeed alignment was allowed by TSs and that the licensee
had adequately evaluated the equipment operation at higher than normal
voltages.
Within the areas inspected, no violations or deviations were identified.
7.
Plant Support (71750)
7.1
7.2
Injured Worker in Containment
On February 15, a rachet wrench fell approximately thirty.feet and
injured a worker in the C RCS loop room.
The worker received a
small cut to the head.
Another worker in containment promptly
reported the injury to the control room.
The Unit 2 SRO
immediately dispatched the First Aid response team.
EMTs provided
initial assessment of the individual's injury and evacuated the
worker from containment.
RP technicians verified the worker was
not contaminated. After exiting containment, the worker
complained of a sore neck.
EMTs reevaluated the individual's
condition and recommended transport via ambulance to a local
hospital for observation.
The worker was promptly evacuated,
treated at the hospital, and returned to work the next day.
The
inspectors observed licensee response to this event and discussed
followup actions with licensee management.
The inspectors
concluded that on-site medical response to the event was 9009 and
that follow-up actions were appropriate.
Containment ALARA
During tours of the Unit 2 containment the inspectors noted that
good planning and training were evident.
RP staff were
consistently challenging workers as to the purpose of their entry,
as well as, providing guidance and instructions to reduce
exposure.
Water shield tanks were observed surrounding the stored
reactor head and lead shielding was used extensively throughout
containment.
Within the areas inspected, no violations or deviations were identified *
- 8.
10
Licensee Event Report Followup (92700}
The inspectors reviewed the LERs listed below and evaluated the adequacy
of the corrective action.
The inspectors' review also included followup
of the licensee's corrective action implementation.
8.1
8.2
(Closed) LER 50-280/93-001, Reactor Trip And Safety Injection Due
To Spurious High Consequence Limiting Safeguards Signal. The LER
reported a reactor trip and safety injection due to a single relay
failure in the safeguard circuitry. The response to the reactor
trip and inunediate corrective actions were discussed in NRC
Inspection Report Nos. 50-280/93-03 and 50-281/93-03.
The inspectors verified,that licensee's commitments contained in
the LER's Actions To Prevent Recurrence section were properly
implemented.
Specifically, the inspectors verified that the NSSS
vendor drawing 113E243 sheet 6, revision 12, included test switch
TS-CLS-lA and 18 contacts 2.
In addition, the inspectors verified
that the event was included in lesson resources that are used in
introductory and requalification training sessions for safety
evaluation preparers and reviewers.
The study to identify single
relays that can fail and cause a reactor trip and the preventive
maintenance practices to be implemented for these relays was
inspected as part of the closeout for LER 50-280/93-002 and is
documented in NRC Inspection Report Nos. 50-280/94-08 and
50-281/94-08.
(Closed) LER 50-281/93-002, Unit 2 Automatic Reactor Trip Due To
Low Steam Generator Water Level Coincident With Steam/Feedwater
Flow Mismatch Resulting From Main Feedwater Pump Trip.
The LER
discussed a reactor trip that resulted from a loss of feedwater
due to an electrical ground in the A MFP inboard motor.
The
reactor trip and inunediate corrective actions were addressed in
NRC Inspection Report Nos. 50-280/93-15 and 50-281/93-15.
In the
LER, the licensee committed to perform a RCE to determine why the
motor failed.
The inspectors reviewed RCE 93-11 that concluded
the failure was most likely due to foreign material that either
was left in the motor when it was rebuilt or fell into the motor
after it was re-installed. The inspectors verified that
O-ECM-1406-01, Main Feedwater Pump Motor Maintenance, revision 3,
step 6.3.1 required FME controls to be initiated. Proper FME
implementation should preclude similar motor failures.
8.3
{Closed) LER 50-281/93-004, Unit 2 Turbine-generator Trip Via
The Loss Of Field Relay.
The turbine-generator/reactor trip
reported in this LER was discussed in NRC Inspection Report Nos.
50-280/93-22 and 50-281/93-22.
In the LER the licensee conmitted
to review voltage regulator performance.
The licensee detennined
that the failure was not similar to previous voltage regulator
failures experienced at the station. However, in meetings with
Westinghouse and the voltage regulator vendor, several items were
identified which could improve the voltage regulators'
11
reliability. The inspectors verified that these items were being
implemented.
The LER also reported a spurious closure of Fire Door 18 and
discussed position indication problems with control rod M-10.
An
engineering review failed to identify the cause of the spurious
fire door closure.
No additional spurious actuations have
occurred. During the present refueling outage, repairs were made
to M-10 position indicating components.
However, success of these
efforts cannot be determined until the position indicator's
performance is observed during a reactor shutdown.
This later
item continues to be tracked by the licensee. Based upon the
completed commitments and the planned actions associated with
M-10, this LER is considered closed.
8.4
(Closed) LER 50-280, 281/93-009, Mechanical Equipment Room #4 Fire
Door Left Blocked Open Due To Personnel Error.
The LER involved a
failure to maintain a fire watch on an open fire door.
The LER
indicated that this event and fire watch responsibilities were
discussed with fire watch qualified personnel and .their
supervision.
The inspectors verified that training on this LER
had been included in fire watch lesson resource NET-9-LP-l, Fire
Watch Training, and in fire watch reverification lesson resource
NECT-9-LP-l, Fire Watch Reverification .
8.5
(Closed) LER 50-280/93-014, Delta Flux Not Logged While Alarm Was
Inoperable Due To Procedural Deficiency.
The LER described a
condition in which a partial failure in the Prodac-250 computer
resulted in the axial flux difference not being logged and
assessed as required by TS.
The need to perform the TS action was
not identified after a computer problem light was received because
the displayed values appeared reasonable.
However, a malfunction
in the computer's integration an4 averaging functions resulted in
the displayed values not being updated and thus invalid.
In the
LER's Action To Prevent Recurrence section, the licensee co11111itted
to revise O-AP-20.02, Loss of the Prodac-250 Computer, to address
partial and complete computer failures.
The inspectors reviewed
O-AP-20.02, revision 2, and confirmed that instructions, as well
- as, entry conditions for partial loss of the Prodac computer were
incorporated into the procedure.
Proper implementation of this
procedure should help preclude similar events from occurring.
In
addition, the inspectors verified that the O-AP-20.02 revision
and this event were incorporated into LORP Training Synopsis
RQ-94-4TS-10 *
. 8.6
(Closed) LER 50-280/93-015, More Than One Individual Rod Position
Indication Channel Per Group Inoperable. The LER reported a loss
of all control and shutdown rod IRPis due to a momentary ground
created when a signal conditioning module with an extension board
was inserted into the instrument racks during calibration. The
LER attributed the fault to the insertion of the extension card
into the modular plug. Subsequent to the LER submittal, the
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L
L
12
ground was attributed to a broken wire on the extension card.
The
card was repaired and no further corrective actions were taken.
The. inspectors concluded that the licensee's response to this
event was appropriate.
Within the areas inspected, no violations or deviations were identified.
9.
Action on Previous Inspection Items (92901, 92903)
9.1
(Closed) URI 50-280, 281/93-15-01, Use Of PRA For Unreviewed
Safety Question Determination.
prepared to justify placing a MFRV on its jack. Both SEs were
based upon engineering judgement that was bolstered by PRA
considerations.
SE 93-155 superseded SE 93-142.
The inspectors
determined that the engineering judgements without the PRA
arguments were sufficient justification to support the
acceptability of SE 93-155.
The URI was opened to explore whether
PRA considerations, as provided in NSAC-125, Guidelines For 10 CFR
50.59 Safety Evaluations, could be used in performing 10 CFR 50.59
reviews. According to cognizant NRR personnel, the staff has not
approved a simple reference to NSAC-125 as an acceptable method of
evaluation against the criteria specified in 10 CFR 50.59.
Acceptable use of PRA and IPEs are currently being examined by the
staff. Since the subject is germane to the industry and not just
this licensee, this item is considered closed.
9.2
(Closed) URI 50-280, 281/93-26-01, EOP Adequacy.
The URI
concerned an interpretation of a WOG standard procedure relating
to when AFW flow could be throttled to limit RCS cooldown after a
reactor trip. After review, the licensee determined that AFW flow
should be throttled so that the RCS temperature would be
maintained at or trending toward 547 degrees F.
AFW flow would
not be throttled below the minimum value specified for
establishing an adequate heat sink.
The inspectors verified that
l-ES-0, Reactor Trip or Safety Injection, revision 15, step 20 and
l-ES-0.1, Reactor Trip Response, revision 12, step 1 contained the
appropriate instructions to limit cooldown by throttling AFW flow.
LORP lesson plan RQ-94-2-LP-DRR, Emergency Operating Procedure
Revisions, was utilized to inform operating personnel of this
change.
9.3
(Closed) IFI 50-280, 281/94-31-02, SRF Overpressurization NOUE -
Control of Work Activities. This item involved control of
licensee work activities which led to a chemical waste treatment
tank rapid overpressurization on November 25, 1994.
The
overpressurization injured one person and resulted in a NOUE.
Radiological contamination was not involved. The licensee
initiated RCE 94-24 and an independent incident review by a
contractor. The RCE detennined that existing procedures for
operating the SRF wet oxidation system did not address the
chemical addition method used on November 25.
In addition,
ineffective management oversight was identified as a contributing
13
cause.
These findings were consistent with those of the
independent contractor incident review. During followup~ the
inspectors experienced difficulty obtaining complete information
through personnel interviews. This difficulty did not alter the
inspectors' course of action or final conclusions. Licensee
management subsequently took action to assure complete and
accurate information was provided.
The licensee initiated several
immediate and long term corrective actions to address the event.
Actions included developing procedures to cover a wider variety of
system operations, safety evaluation of system operation, and
personnel actions. The inspectors concluded that these actions
were .appropriate to preclude recurrence.
Within the areas inspected, no violations or deviations were identified.
10.
Exit Interview
The inspection scope and findings were summarized on March 7, 1995, with
those persons indicated in paragraph 1.
The inspectors described the
areas inspected and discussed in detail the inspection results addressed
in the Summary section and those listed below.
Item Number
Status
DescrigtionL(Paragragh No.}
NCV 50-281/95-05-01
Closed
Failure to Monitor Load Cell
(paragraph 3.1.3).
LER 50-280/93-001
Closed
Reactor Trip And Safety
Injection Due To Spurious High
Consequence Limiting
Safeguards Signal
(paragraph 8.1).
LER 50-281/93-002
Closed
Unit 2 Automatic Reactor Trip
Due To Low Steam Generator
Water Level Coincident With
Steam/Feedwater Flow Mismatch
Resulting From Main Feedwater
Pump Trip (paragraph 8.2).
LER 50-281/93-004
Closed
Unit 2 Turbine-generator Trip
Via The Loss Of Field Relay
(paragraph 8.3).
LER 50-280, 281/93-009
Closed
Mechanical Equipment Room #4
Fire Door Left Blocked Open
Due To Personnel Error
(paragraph 8.4).
' L
~ *
14
Item Number
Status
Description/(Paragraph No.)
LER 50-280/93-014
Closed
Delta Flux Not Logged While
Alarm Was Inoperable Due To
Procedural Deficiency
(paragraph 8.5).
LER 50-280/93-015
Closed
More Than One Individual Rod
Position Indication Channel
Per Group Inoperable
(paragraph 8.6).
URI 50-280, 281/93-15-01
Closed
Use Of PRA For Unreviewed
Safety Question Determination
(paragraph 9.1).
URI 50-280, 281/93-26-01
Closed
EOP Adeq~acy (paragraph 9.2).
IFI 50-280, 281/94-31-02
Closed
SRF Overpressurization NOUE -
Control of Work Activities
(paragraph 9.3).
Proprietary information is not contained in this report . Dissenting
comments were not *received from the licensee.
11.
Index of Acronyms and Initialisms
CFR
CV
IFI
!RPI
Kv
LER
LORP
HFRV
NFA
NRC
NSAC
AS LOW AS REASONABLY ACHIEVABLE
CODE OF FEDERAL REGULATIONS
CONTAINMENT VESSEL
EMERGENCY MEDICAL TECHNICIANS
EMERGENCY OPERATING PROCEDURE
INSPECTION FOLLOWUP ITEM
INDIVIDUAL PLANT EXAMINATIONS
INDIVIDUAL ROD POSITION INDICATION
KILOVOLTS
LICENSEE EVENT REPORT
LICENSED OPERATOR REQUALIFICATION PROGRAM
MAIN FEEDWATER PUMP
MAIN FEEDWATER REGULATING VALVE
MAIN FEEDWATER
NON-CITED VIOLATION
NUCLEAR FUEL ANALYSIS
NOTICE OF UNUSUAL EVENT
NUCLEAR REGULATORY COMMISSION
NUCLEAR REACTOR REGULATION
NUCLEAR SAFETY ANALYSIS CENTER
NUCLEAR STEAM SUPPLY SUPPLIER
" *
RV
SRF
TS
V
15
ROOT CAUSE EVALUATION
REFUELING OUTAGE
RADIATION PROTECTION
RESERVE STATION SERVICE TRANSFORMER
REACTOR VESSEL
REFUELING WATER STORAGE TANK
SAFETY EVALUATION
SAFETY INJECTION
SURRY RADWASTE FACILITY
SENIOR REACTOR OPERATOR
TECHNICAL SPECIFICATION
UNRESOLVED ITEM
VOLTS
WESTINGHOUSE OWNERS GROUP